ML20209C194

Typed,Final TS Pages for LAR 115.Summary Description of Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages

ML20209C194
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Site:	Beaver Valley
Issue date:	06/28/1999
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ML20209C193	List: L-99-103, Supplemental Application for Amend to License NPF-73, Including Typed,Final TS Pages for LAR 115.Summary Description of Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages ML20209C194
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NUDOCS 9907090196
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ATTACHMENT 115 Technical Specification Paces i

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s 9907090196 990628 PDR ADOCK 05000412 PM p

ATTACHMENT TO LICENSE AMENDMENT NO.

FACILITY OPERATING LICENSE NO. NPF-73 DOCKET NO. 50-412 Replace the following pages of Appendix A, Technical Specifications, with .the enclosed pages as indicated. The revised pages are

-identified by amendment number and contain vertical lines indicating the areas of change.

Remove Insert 1-4 1-4 1-5 1-5 1-6 1-6 1 1-7 3/4 4-27 3/4 4-27 3/4 4-28 3/4 4-28 3/4 4-29 3/4 4-29 B 3/4 4-3b B 3/4 4-3b B 3/4 4-4g B 3/4 4-4g B'3/4 4-4h B 3/4 4-4h B 3/4 4-5 B 3/4 4-5 B 3/4 4-6 B 3/4 4-6 B 3/4 7-2j B 3/4 7-2j s

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NPF DEFINITIONS 1.15 THROUGH 1.17'(DELETED)'

OUADRANT POWER TILT RATIO (OPTR) l'.18 QPTR shall be the ratio of the maximum upper excore detector calibrated output to 'the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated ~ output to the: average of the lower. excore detector calibrated outputs, whichever is greater.

DOSE EOUIVALENT I-131 1.19 . DOSE EQUIVALENT ' I-131.' shall be that concentration of I-131 (microcuries/ gram)' that .alone would produce the same thyroid dose as the quantity and isotopic mixture of'I-131, I-132, I-133, I-134, and I-135 actually present. The DOSE EQUIVALENT I-131 is calculated with the following equation:-

• ~+

Cg_131 DJ. = C g_131 + + +

170 6- 1000- 34 l

Where "C" is the concentration, in microcuries/ gram of the iodine isotopes. This equation is based on dose conversion factors derived q from ICRP-30. 1 I

STAGGERED TEST BASIS 1.20 A STAGGERED TEST' BASIS shall consist of:

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a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval-into n equal subintervals;

b. The testing of one -(1) system, subsystem, train or other designated component at the.beginning of each subinterval.

FREOUENCY NOTATION i

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1.21- The FREQUENCY NOTATION specified for the performance of '

Surveillance Requirements shall correspond to the intervals defined

'in Table 1.2. j REACTOR TRIP SYSTEM RESPONSE TIME 1.22 -The REACTOR TRIP SYSTEM - RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.

BEAVER VALLEY - UNIT 2 1-4 Amendment No.

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NPF-73 DEFINITIONS ENGINEERED SAFETY FEATURE RESPONSE TIME 1.23 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from' when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required valras, etc.). . Times shall include diesel generator starting and sequence loading delays where applicable'.

AXIAL FLUX DIFFERENCE 1.24 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two-section excore neutron detector.

PHYSICS TESTS 1.25 PHYSICS' TESTS shall.be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR,

2) authorized-under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

E - AVERAGE DISINTEGRATION ENERGY 1.26 E. shall be the average sum (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15-minutes, making up at least 95% of the total non-iodine activity in the coolant.

SOURCE CHECK 1.27 A SOURCE. CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

PROCESS CONTROL PROGRAM 1.28 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas,' sampling,. analyses, test, and determinations to be made to ensure . that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes-will be accomplished in such a way as to assure compliance with 10 CFR~ Parts 20, 61, and 71, State regulations,' burial ground requirements, and other requirements governing the disposal of solid radioa'ctive waste.

1.29 DELETED BEAVER VALLEY - UNIT 2 1-5 Amendment No.

NPF-73 DEFINITIONS OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.30 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs ,

required by Section 6.8.6 and (2) descriptions of the information l that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications in the Administrative Control section.

GASEOUS RADWASTE TREATMENT SYSTEM 1.31 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting Primary Coolant System offgases from the primary system and providing for delay or holdup for the purpose of reducing the total l radioactivity prior to release to the environment.

VENTILATION EXHAUST TREATMENT SYSTEM  ;

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1.32 VENTILATION EXHAUST TREATMENT SYSTEM is any system designed f and installed to reduce gaseous radioiodine or radioactive material  !

in particulate form in effluents by passing ventilation or vent l exhaust gases through charcoal adsorbers and/or HEPA filters for the l purpose of removing lodines or particulates from the gaseous exhaust 1 stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESP) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

PURGE-PURGING 1.33 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating conditions, in such a manner that replacement air or gas is required to purify the confinement.

VENTING 1.34 VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating conditions, in such a manner that replacement air or gas is not provided or required during VENTING.

Vent, used in system names, does not imply a VENTING process.

BEAVER VALLEY - UNIT 2 1-6 Amendment No.

'NPF-73 DEFINITIONS MAJOR CHANGES 1.35 MAJOR CHANGES to radioactive waste systems (liquid, gaseous and solid), as addressed in the PROCESS CONTROL PROGRAM, shall include the following:

1) MAJOR CHANGES in process equipment, components, structures, and effluent monitoring instrumentation from those described in the Final Safety Analysis Report (FSAR) or the Hazards Summary Report and evaluated in the staff's Safety Evaluation Report (SER) (e.g., deletion of evaporators 'and installation of demineralizers; use of fluidized bed calciner/ incineration in place of cement solidification systems);

2) MAJOR CHANGES in the design of radwaste treatment systems (liquid, gaseous, and solid) that could significantly increase the quantities or activity of effluents released or volumes of solid waste stored or shipped offsite from those previously considered in the FSAR and SER (e.g.,

use of asphalt system in place of cement);

3)- Changes in system design which may invalidate the accident analysis as described in the SER (e.g., changes in tank. capacity that would alter the curies released);

and

4) Changes'in system design that could potentially result in a significant increase in occupational exposure of operating personnel (e.g.,- use of - temporary equipment i without adequate shielding provisions).

MEMBER (S) OF THE PUBLIC 1.36 MEMBER (S) OF THE PUBLIC means any individual except when that individual is receiving an occupational dose.

CORE OPERATING LIMITS REPORT 1.37 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.5. .

Plant operation within these operating limits l is addressed in individual specifications.

BEAVER VALLEY --UNIT 2 1-7 Amendment No.

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NPF-73 REACTOR COOLANT SYSTEM 3/4.4.8' SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION L3.4.8 The spec'ific activity of the reactor coolant shall be limited

.to:'

a.. $ 0.35 pCi/ gram DOSE EQUIVALENT I-131, and l

b. s 100/E pC1/ gram.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION: ,

MODES 1, 2 and 3*:

a. With the specific activity of the primary coolant > 0.35 l pCi/ gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3. 4-1, be in HOT STANDBY with Tavg

< 500'F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With the specific activity of the primary coolant > 100/E pCi/ gram, be in HOT STANDBY with Tavg < 500 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

MODES 1, 2, 3, 4, and 5 l

a. With the specific activity of the primary coolant > 0.35 l pCi/ gram DOSE EQUIVALENT I-131 or > 100/E pCi/ gram, perform

! the sampling analysis requirement of item 4a of Table

4. 4 until the specific activity of the primary' coolant l is restored to within its limits.

SURVEILLANCE REQUIREMENTS , 4.4.8 The specific activity of the primary coolant shall be l

! determined to be within the performance limits of the sampling and analysis program of Table 4.4-12.

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• With Tavg A 500*F.

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BEAVER VALLEY - UNIT 2 3/4 4-27 Amendment No.

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NPF-73 TABLE 4.4-12 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT MINIMUM MODES IN WHICH AND ANALYSIS FREQUENCY SURVEILLANCE REOUIRED

1. Gross Activity 3 times per 7 days 1, 2, 3, 4 Determination with a maximum time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> between samples.

2. Isotopic Analysis 1 per 14 days 1

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for DOSE EQUIVALENT I-131 Concentration

3. Radiochemical -for 1 per 6 months 1 l E Determination

4. Isotopic Analysis a) Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 1#,2#,3#,4#,5#

for Iodine whenever the including I-131 specific activity I-133, and I-135 exceeds 0.35 l pCi/ gram DOSE EQUIVALENT I-131 or 100/E pCi/ gram, and b) One sample between 1, 2, 3 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a THERMAL POWER 4 change exceeding 15 percent of the RATED THERMAL POWER within a 1-hour period.

1. Until the specific activity of the primary coolant system is restored to within its limits.

BEAVER VALLEY - UNIT 2 3/4 4-28 Amendment No.

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NPF-73 1

2 250 \ j 20

,5 t-2 l J. J D

5: 200 8

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\ UNACCEPTABLE OPERATION

[E D

E 150 5

8 O

E 100 E

s E ACCEPTABLE OPERATION a

% 50 (

5 8

8 8 l 0 l 20 30 40 50 60 70 80 90 100 l PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUlVALENT l-131 Primary Coolant Specific Activity Limit Versus l Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 0.35 pCi/ gram Dose Equivalent 1-131 BEAVER VALLEY - UNIT 2 - 3/4 4-29 Amendment No.

NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS (Continued) where Vor represents the allowance for degradation growth between inspections and.Vuos represents the allowance for potential sources of error in the measurement of . the bobbin coil voltage'.- Further discussion ' of . the assumptions necessary to determine the voltage repair limit are discussed in GL 95-05.

Safety analyses were performed pursuant to Generic Letter 9s-05 to

~ determine the maximum MSLB-induced primary-to-secondary leak rate <

that could occur without offsite doses excceding a small fraction of 10 CFR.100 (concurrent iodine spike), 10 CFR 100 (pre-accident iodine spike), and without control room doses exceeding GDC-19. The current value of ~ the maximum MSLB-induced leak rate and a summary of the analyses are provided in Section 15.1.5.of the UFSAR.

The mid-cycle equation in SR 4.4.5.4.a.10.d should only be used during unplanned inspections in which eddy current data is acquired for indications at the tube support plates.

SR 4.4.5.5 implements several reporting requirements recommended by GL 95-05 for situations.which the NRC wants to be notified prior to returning the SGs to service. For the purposes of this reporting requirement, LEAKAGE and conditional burst probability can be calculated based on the as-found voltage distribution rather than the projected end-of-cycle (EOC) . voltage distribution (refer to GL 95-05 for more information) when it is not practical to complete these 4 calculations using the projected EOC voltage distributions prior to returning the SGs to service. Note that if LEAKAGE and conditional burst probability were calculated using the measured EOC voltage distribution for the purposes of addressing the GL section 6.a.1-and 6.a.3 reporting criteria, then the results of the projected -EOC voltage distribution should be provided per the GL section . 6.b (c) criteria.

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pursuant to Specification 6.6 prior to resumption of plant operation. Such cases will be' considered by the Commission on <!

a case-by-case basis and may result in a requirement for analysis, laboratory . examinations, - tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

BEAVER' VALLEY - UNIT 2 B 3/4 4-3b Amendment No.

NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

LCO (Continued) component which can be isolated from the balance of the Reactor Coolant System, plant operation may continue provided the leaking component is promptly isolated from the Reactor Coolant System since isolation removes the source of potential failure.

b. Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipmcat can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary,

c. Primarv-to-Secondary LEAKAGE throuch Any One SG Operating experience at PWR plants has shown that sudden increases in leak rate are often precuraors to larger tube failures. Maintaining an operating LEAKAGE limit of 150 esd per steam generator will minimize the potential for a large LEAKAGE event at power. This operating LEAKAGE limit is more restrictive than the operating LEAKAGE limit in standardized technical specifications. This provides additional margin to accommodate a tube flaw which might grow at a greater than expected rate or unexpectedly extend outside the thickness of the tube support plate. This reduced LEAKAGE limit, in conjunction with a leak rate monitoring program, provides additional assurance that this precursor LEAKAGE will be detected and the plant shut down in a timely manner.

BEAVER VALLEY - UNIT 2 B 3/4 4-4g Amendment No.

e NPF-73

' REACT 6R COOLANT SYSTEM i

BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Ccatinued)

LCO (Continued)

d. IdRD*13Jied LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of identified LEAKAGE and is well within the capability of the RCS Makeup System. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressere boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE).

Violation of this LCO could result in continued degradation of a component or system.

APPLICABILITY In' MODES 1, 2, 3, and 4, the potential for RCPB LEAKAGE is greatest

'when the RCS is pressurized.

In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.

LCO 3.4.6.2, "RCS Pressure Isolation Valve (PIV)," measures LEAKAGE through each individual PIV and can impact this LCO. Of the two PIVs in series in each isolated line, I.EAKAGE measured through one PIV does not result in RCS LEAKAGE when the other is leak tight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE.

BEAVER VALLEY - UNIT 2 B 3/4 4-4h Amendment No.

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NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant Systam leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant.

The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent. Corrosion studies show that operation uay be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.

The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

The sorveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

l 3/4.4.8 SPECIFIC ACTIVITY The primary coolant specific activity is limited in order to maintain c,f f site and control room operator doses associated with postulated accidents within applicable requirements. Specifically, the 0.35 pCi/gm DOSE EQUIVALENT I-131 limit ensures that the offsite dose does not excead a small fraction of 10 CFR Part 100 guidelines and that control room operator thyroid dose does not exceed GDC-19 in the event of primary-to-secondary leakage induced by a main steamline break.

The ACTION statement permitting 'JER OPERATION to continue for limited time periods with the prmary coolant's specific activity

> 0.35 pCi/ gram DOSE EQUIVALENT I-131, but within the allowable limit l shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may ot=ur following changes in THERMAL POWER.

Operation with specific activity levels exceeding 0.35 pCi/ gram DOSE l EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limits shown on Figure 3.4-1 must be restricted to ensure that assumptions made in the UFSAR accident analyses are not exceeded.

Reducing Tavg to < 500*F minimizes the release of activity should a l steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam BEAVER VALLEY - UNIT 2 B 3/4 4-5 Amendment No.

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NPF-73 REACTOR COOLANT SYSTEM BASES M4.4.8 SPECIFIC ACTIVITY (Continued) relief valves. This action also reduces the pressure differential across the steam generator tubes reducing the probability and magnitude of main steamline break accident induced primary-to-secondary leakage. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

3/4.4.9 PRESSURE / TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various catagories of load cycles used for design purposes are provided in Section 3.9 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent eith the design assumptions and satisfy the stress limits for cyclic operation.

l During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. These thermal-induced compressive i stresses tend to alleviate the tensile stresses induced by the internal pressure. Therefore, a pressure-temperature curve based on steady state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the I inner wall of the vessel is treated as the governing location. l The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients established during heatup produce tensile stresses at the outer wall of the vessel. These stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup .

ramp; therefore, a lower bound curve similar to that described for l the heatup of the inner wall cannot be definad. Subsequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.

BEAVER VALLEY - UNIT 2 B 3/4 4-6 Amendment No.

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-NPF-73 PLANT

• SYSTEMS BASES 1

3/4.7.1.3 PRIMARY PLANT DEMINERALIZED WATER (PPDW)

I The ' OPERABILITY of the PPDW storage tank with the minimum water I volume ensures that sufficient water is available to maintain the RCS

.at HOT -STANDBY conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam discharge to atmosphere.

3'/ 4 . 7 .1.' 4 ACTIVITY The limitations on secondary ~ system specific activity ensure that steam releases to . the . environment will not be significant contributors to radioactivity releases resulting from analyzed accidents. Many of the analyzed accidents assume that a loss of 1 l auxiliary AC power occurs, making the main condenser unavailable for plant cooldown, and making it necessary to dump steam to the environment via SG atmospheric dump valves. ' Maintaining secondary system' specific activity within.the limits ensures that these releases, in conjunction with other releases associated with the accident, will be within applicable dose criteria.

' BEAVER' VALLEY - UNIT 2 B 3/4 7-2j Amendment No.

4 ATTACHMENT B-1 4

Change Summary l

Typed Paces ' Editorial Chances l'-4 No additional changes are included. I 1-5 1-6 i

1-7 In Definition 1.37, for the Core Operating Limits Report, reference to Specification "6.9.1.12" is replaced with Specification "6.9.5" since i the specification number was changed in Amendment 97.

3/4 4-271 In Specification 3.4.8.b, a period was added following " Ci/ gram" and also in the Applicability, following " Mode 5" to be consistent with other specifications.

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3/4 4-28 In Table 4.4-12 in the column titled ' Modes in Which Surveillance l Required', the comma has been removed following "1" for Items 2 and 3. The comma is not needed since no other modes are applicable.

3/4.4-29 No additional changes are included.

B 3/4 4-3b ,

These pages have been updated to include changes incorporated in B 3/4 4-4g , License Amendment Request (LAR) 109. The amendment for B 3/4 4-4h ' LAR 109 should be issued prior to or coincident with the amendment for LAR 115 to avoid confusion when implementing this amendment.

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l . B 3/4 4-5 In the last sentence of the first paragraph that was added under 3/4.4.8 j l Specific Activity, " steam line" has been replaced with "steamline" to j i be consistent with current convention.

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I B 3/4 4-6 In the third line of the first sentence on this page, " steam line" has been replaced with "steamline" to be consistent with current convention.

1 I B 3/4 7-2j No additional changes are included.

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