ML20031H379

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Testimony of G Martin & Wf Malec Re Mccorkle Contention 17 on Bypass Leakage.Aslb Statement Correct That No Difference Exists in % by Weight & % by Vol
ML20031H379
Person / Time
Site: Allens Creek File:Houston Lighting and Power Company icon.png
Issue date: 09/18/1981
From: Malec W, Martin G
EBASCO SERVICES, INC., HOUSTON LIGHTING & POWER CO.
To:
Shared Package
ML20031H319 List:
References
NUDOCS 8110270428
Download: ML20031H379 (4)
v • d • e


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Ssptrmber 18, 1981 4

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UNITED ST$TES OF AMERICA l'

NUCLEAR.REQULATORY COMMISSION

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,2 BEFORE THE ATOMIC?3AFETY AND LICENSING BOARD 3.

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HOUSTON LIGHTING & POWpd COMPANY S

Docket No. 50-466 51 s,

3 (Allens Creek Nucleah Gr-perating S

Station, Unit 1)

S 7b DIRECT TESTIMONY OF GUY MARTIN, JR. AND WALTER F. MALEC S

REGARDING McCORKLE CONTrNTION-NO. 17 - BYPASS LEAKAGE 9

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Mr. Martin and Mr. Malec, have you reviewed your 0.,

affidavit on Meccrkle contention'No. 17, which affidavit is 11 N.

i attacned hereto as Attachment SM/WFM-l?

12u Yes..

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sQ i-Are the sta[yrnents contained in the affidavit

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e15 s: 'Q except for the7 changes described in the errata y

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  • attached hereto as Attachment GM/WFM-2, 17 j

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Mr. Martin, have yo'u reviewed that portion of the 18 4

Board'O Order of Sept. ember 1, 1981, wherein the Board 19 2

A h 5 y calculatedthe} amount.ofunfilteredleakaga (0.0195%) to be 2,0

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approximately,40% of the 0.5% total leakage?

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Y ea, I have.

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Would you please address the questions raised by

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,the Board at;pages 4 andub,,.of the Order?

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1 order reveals that an arithmetical error has been made.

The 2

bypass leakage is 0.0195% of the containment volume per day r approximately 4% of the 0.5% containment leakage rate 3

value. The Board's statement concerning the calculation 4

methodology used to arrive at the bypass leakage value is correct.

however, it should be noted that the presently calculated bypass leakage value of 0.0195% of the containment volume, if it were to occur, would result in a thyroid dose 8

value equal to one-half of the 10 CFR Part 100 thyroid dose 9

limit.

As s'tated in the Supplement No. 2 of the Staff 10 Safety Evaluation Report, the atmospheric dicpersion factor 11 at the exclusion zone boundary has decreased.

However, the 12 bypass fraction of 0.0195% is based on a previously calculated j

13 atmospheric dispersion factor which is 67% higher than the 14 dispersion factor which would have been calctilated using 15 current NRC guidance and site meteorological deta.

Conse-16 quently, offsite doses would be significantly lower than 17 previously determined if they were calculated using this 18 bypass fraction in conjunction with the current NRC Staff 19 atmospheric dispersion factors.

At the Operating License stage the bypass leakage value will be recalculated to 29 reflect the latest NRC methodology and site meteorological 21 date to calculate the site-specific atmospheric dispersion 22 fa t rs.

3 The Board's mention of the containment leak rate 24

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O which was specified as both'a percentage of weight and volume 1

denotes that th. presentation of this value as a function of 3

these tw'o parameters has caused a degree cf confusion which warrants some clarification.

4 In the calculation of the offsite radiological 5

doses to show compliance with the siting criteria of 10 CFR 6

Part 100, the containment is assumed to leak'at a constant 7

leak rate of 0.5% of its volume per day.

From a dose 8

calculation standpoint, the racionuclides, uniformly mixed 9

in the containment atmosphere, are assumed to leave the con-10 tainment at this constant leakage rate regardless of the flow 11 rate of carrier air in which they are assumed to be mixec.

12 The maximum containment airborne concentration of these radio-13 nuclides will occur at standard temperature and pressure (STP) 14 conditions.

Therefore, the air leakage expressed in terms 15 of a fraction of the containment air volume at STP conditions 16 will have the same radionuclide concentration and hence will 17 be selected as the technical specification value to be met, in 18 testing, in order to remain within the dose criteria of 10 CPR Part 100.

The leakage rate can be expressed as a 19 Percentage of weight per a unit of time by converting volume 20 to weight.

Under test conditions, the containment will be 21 pressurized, the leak rate measured and compared to this technical specification value.

The Board's statement is correct in that there is no difference in percent by weight 24 I

1 and percent by volume no matter how it is expressed, since, ultimately, the actually measured quantity will be either a 2

mass or a volume of air per a unit of time.

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Attachmsnt GM-1

_. UNITE _D_ STATES _OF AMERICA _._.

,j, NUCLEAR REGULATORY COMMISSION

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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD i

l In the Matter of

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r HOUSTON LIGHTING & POWER

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Docket No. 50-466 COMPANY

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(Allens Creek Nuclear

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Generating Station, Unit

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AFFIDAVIT OF GUY MARTIN, JR.

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State of New Jersey

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County of Bergen

'I 1, Gu:r Martin, Jr., Supervising Radiological Assessment Engineer, Allens Creek Pro j ect, for Ebasco Services Incorporated, of lawful age, being first duly sworn,

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upon my cath certify that I have reviewed and am thoroughly familiar with the statements contained in the attached affidavit addressing intervenor Brenda f

McCorkle's Contention 17 regarding filtration system leakage. All statements contained therein,which relate to Ebasco Services Incorporated scope of supply for the Allens Creek Nuclear Generating Station, are true and correct to the

'7 best of my knowledge and belief.

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Subscribed and sworn to before me this # j(I day f N

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  • p.D CAROL 'A. OPITEN t,

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7 NOTARY PU90C E NEW JER$FY f

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MY COMMISSION DFtRES SEM. 18, 1983

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UNITED STATES OF AMERICA g

NUCLEAR REGULATORY COMMISSION I

BEFORE THE AT0 HIC SAFETY AND LICENSING BOARD l

In the Matter of

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HOUSTON LIGHTING & POWER

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Docket No. 50-466 COMPANY

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(Allens Creek Nuclear

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Generating Station, Unit

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AFFIDAVIT OF WALTER F MALEC g-State of New Jersey County of Bergen

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I, Walter F Malec, Supervising Mechanical Nuclear Engineer, Allens Creek

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Project, for Ebasco Services Incorporated, of lawful age, being first duly sworn, upon my oath certify that I have reviewed and am thoroughly familiar

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with the statements contained in the attached affidavit addrissing intervenor Brenda McCorkle's Contention 17 regarding filtration system 1'akage and that all statements contained therein are true and correct to the itst of my knowledge and belief.

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Subscr? bed and sworn to before me thfs

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NOTMY PUBUC OF NEW jrpm

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CAROL A. CPITENOK FAY CD:,rd;SSION EXP!RES SEPT. 13,1983 lI V

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Attachmsnt GM-1 l

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of S

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HOUSTON LIGHTING & POWER S

COMPANY S

Docket No. 50-466 S

(Allens Creek Nuclear S

Generating Station, Unit S

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No. 1)

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AF'IDAVIT.0F GUY MARTIN, JR.

AND WALTER F. MALEC k

l My name is Guy Martin, Jr.

My business address is Two l

World Trade Center, New York, N. Y.

I am the Supervising Radio-

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logical Assessment Engineer for the Allens Creek Project employed c

I by Ebasco Services Incorporated.

The statement of my background i.

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and qualifications is attached as Exhibit I to this testimony.

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My name is Walter F. Malec.

My business address is 160 b

Chubb Avenue, Lyndhurst, N.

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I am the Supervising Mechanical u

Nuclear Engineer for the Allens Creek Project employed by Ebasco Services Incorporated.

The statement of my background and l (~

qualifications is attachedias Exhibit II to this testimony.

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This affidavit addresses the issues raised in McCorkle Contention No. 17.

The contention states that the Allens Creek containment as designed will allow 20 percent of the containment leakage to bypass the filtration systems.

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Introduction i

The Allens Creek containment consists of a free-standing I

steel shell 1 1/2 to 1 3/4 inches thick which encloses the reactor vessel holding the reactor fuel.

The containment is designed to protect the public from the release of radioactive

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fission products by providing a leak-tight barrier.

However, for practical purposes, the containment must be penetrated by piping and other openings.

Although these penetrations are sealed by some means such as redundant valving, a certain quantity of leakage is inevitable.

NRC regulations (10 CFR, Part 50, j {

Appendix J) limit the quantity of leakage allowed.

4 II.

Containment Leakage Expected for Allens Creek f

4 The Containment vessel is a seismic Category.I steel shell designed to confine the radioactive materials, gases under pressures and temperatures associated with a loss-o'f-coolant r

accident and all other abnormal operating conditions.

The design leak rate will.be 0.5 percent by weight of the contained atmosphere per day at calculated peak pressure.

The Containment Vessel will be designed to contain any leakage from the drywell and the

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noncondensable gases from teactor vessel blowdown by the safety / relief valves or from the rupture of the largest pipe inside the drywell.

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To determine the type of leakage which can be expected,

' 1 a list of all potential leakage paths through containment penetrations t

was compiled (Table 6.2-12a of the Preliminary Safety Analysis s _,

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I Report).

This list is reproduced as Exhibit A.

From this list, only six penetrations constitute potential unfiltered leakage I

paths.

These six penetrations are listed in Table 6.2-13 of the PSAR and the table is reproduced as Exhibit B.

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In arriving at the list contained in Exhibit B, an

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evaluation was made of all lines which penetrate the containment to determine the number and types of barriers to bypass leakage l

provided for each line.

The types of bypass leakage barriers i

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considered were as follows:

(a)

Isolation valve outside containment.

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(b)

Isolation valve inside containment.

(c)

Closed Categcry I piping system inside containment.

2 (d)

Closed Category I piping system outside

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containment.

(e)

Water seal in line, c-I (f)

Line beyond isolation valve outside contain-

.I ment vented to annulus for filtration by the Standby 7

4g Gas Treatment System (SGTS).

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(g)

Line terminates outnide containment in filtered

_I ECCS Area of Auxiliary Building.

Leakage barriers of types (c) through (g) effectively eliminate l

any bypass leakage.

Leakage barriers of types (a) or (b) i limit but do net eliminate bypass leakage.

Therefore, lines L

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containing any of ens bypnco leakaga barriera (c) through (g)

I were not considered as potential bypass leaksge paths.

Lines containing only types (a) or (b) were included in Exhibit B as l

potential unfiltered leakage paths.

g III.

Unfiltered Leakage The amount of containment leakage allowed in the I

Technical Specifications will be significantly less than that which would produce total off-site doses equal to the l

10 CFR 100 limits.

The contributors to this total leakage include p

g the Standby Gas Treatment System releases, leakage to the con-

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trolled ventilation ECCS area of the Auxiliary Building and all I

unfiltered bypass leakage.

The actual value of the bypass leakage r

technical specification will be determined as a result of LOCA 1.

r dose calculations performed when the FSAR 16 prepared for submittal.

Hovever, a value of.0195 percent / day of the containment volume

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is the present best estimate of the maximum total unfilterad bypass leakage based on preliminary LOCA dose calculations.

These dose calculations are provided in detail in Section 15 and Appendlx 15.A of the PSAR.

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IV.

Tests and Inspections f~

In order to assure that the containment will maintain its expected level of leak-tightness, Applicant will conduct a leak testing program in accordance with f

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The fraction of total containment leak rate technical specification which will be released via potential bypass

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leakage lines is quoted at PSAR, p. 15.A-4b as 2.9 x 10-2, This number is a typographical error.

The correct value is g'.

3.9 x 10-2, L

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l Appendix J of 10 CFR 50.

As required by Appendix J, three ypes of tests will be performed:

I Type A - This test will measure the primary reactor containment overall integrated leakage rate.

It will be conducted after the containment is completed and ready

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for operation and again about once every three and one-third years thereafter.

In addition, any major modification

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or replacement of components of the primary reactor containment performed after the initial leak rate tect shall be followed by either a Type A test or a Type B

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test of the area affected by the modification.

Type B - Appendix J defines these tests as those:

intended to detect local leaks and to measure leakage across each pressure-containing or leakage-limiting boundary for the following prie,r;y reactor containment penetrat1Gns:

l.

Containment penetrations 7

whose design incorporates resilient f

seals, gaskets, or sealant compounds, l-piping penetrations fitted with expansion bellows, and electricci penetrations fitted with flexibic metal seal assemblies.

i 2.

Air Inck door seals, including decr operating mechanism penetrations which are part of the containment 7

pressure boundary.

3.

Doors with resilient seals or i

gaskets except for seal-welded doors.

4.

Components other than those u

listed.above which must meet the acceptance criteria in III.B.3 of Appendix J.

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Excspt for containment air locks, Type B tests will t I

conducted during each reactor shutdown for major fuel reloading but in no case at intervals greater than two years.

The seals of the personnel air locks will be t-tested after each opening or, if left unopened, at an interval not to exceed one year.

l Type C - Type C tests are those intended to measure containment isolation valve leakage rates.

The contain-p ment isolation valves included are those that:

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1.

Provide a direct connection between the inside and outside atmospheres of the primary reactor containment under p

normal operation, such as purge and ventila-tion, vacuum, relief, and instrument

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valves; b

2.

Are required to close auto-

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matically upon receipt of a containment isolation signal in response to controls intended to effect containment isolation; r

3.

Are required to operate intermit-tently under post-accident conditions; and I

4.

Are in main steam and feedwater piping and other systems which penetrate containment of direct-cycle boiling water r

power reactors.

Type C tests shall be performed for isolation valves during each reactor shutdown for major refueling.

1 V.

Conclusion The Allens Creek containment will be designed to

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limit leakage to 0.5 percent by weight of the containment atmosphere per day at calculated peak pressure.

Applicant has

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I calculated that, un' der loss of coolant accident conditions, a maximum of.0195 percent per day of containment volume may escape I

via the potential bypass leakage lines and that the resulting i

doses will not exceed the limits of 10 CFR Part 100.

Hence, r

Intervenor's claim that 20 percent of the containment leakage I

will bypass filtration systems does not reflect the present c

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plant dWern and the updated bypass leakage fraction calculations 4

contained in PSAR, Section 15 and Appendix 15.A.

Finally, the i

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projected containment integrity will be assured by performing the leak-rate tests called for by 10 CFR, Appendix J.

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EXHIBIT A I

EVALUATION OF POTENTIAL BYPASS LEAKAGE Ft)R CONTAINMENT l

PENEIRATIONS Line Bypass Cons,1dered Sire Leakage Potential g

System Service

- (in. )

Barriers

  • Bypass Path I

Main Steam Lines 26 A, B. H No A, B, C, and D I

Feedwater A and B 20 A, B, E No RHR Pump A, B, and 24 A,D,E,G No l

C Suction from Sup-pression Pool i

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'g RHR Shutdown Suction 20 A,B,D,E,G No From Recirculation Loop RHR Return A and B 12 A,B,D,E,G No

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to Recirculation

,.i Loop 1.

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RHR A, B, and C 12 A,B,D,E,G No LPCI o

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p RHR A, $, and C 18 A,D,E,G No l

Pu=p Test Lines to l

Suppression Pool HPCS Pu=p Suc" ion 24 A,D,E,G No L

fr m Suppression Pool i

HPCS Pu=p Discharge 12 4, B, D, E.

No f

HPCS Test Line to 12 A, 3. E, G No Suppression Pool

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HPCS Mini =u= Flow 4

A, D, E, G No L.

Line LPCS Pu=p Suction from~

24 A,D,E,G No Suppression Pool

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LPCS Pu=p Discharge 12 A,B,D,E,G No to Pressure Vessel C

LPCS Test Line A,D,E,G No I

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EXHIBIT A 1

Line Bypass Considered Size Leakage Potential l

System Service fin.)

Barriers

  • Bveass Path f

Steau Supply the RCIC 10 A, B, D No l

Turbine and RHR Heat Exchanger g

RCIC and RHR to 6

A,B,D,E No Head Spray g

RCIC Pu=p Suction from 6

A,D,E No Suppression Peel rl P.CIC Turbine EAnaust 12 A, D No I

to.cuppression Pool r

RCIC Pu=p Discharge 2

A,D,E No

!'l Minimum Flow Bypass l

RCIC Vacuum Pump 2

A, G No

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Discharge r

CRD Pu=p Discharge 2

A,B,E No Station Air Supply 2

A, B Yes j-Instru=ent Air 2

A, B Yes Sepply r

Reactor Building 14 A,B,E No I

Closed Cooling Water Supply L

Reactor Building 14 A, B, E No Closed Cooling i

j Water Return L

Reactor Water Clean-4 A, B, E No up to Condenser and

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Radwaste f

j Reactor Water Clean-4 A, B, E No

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up Backwash "ransfer Pu=p Discharge Main Staam Drains 3

A,B,E No b

to Condenser E

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EXHIBIT A

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Line Bypass Considered I

Size Leakage Poten tial System Service

_(in.)

Barriers *

'ivoass Path, I

LPCS Mini =um Flow 4

A D,E,C No Liite l

RHR Pt=p Minimum 4

A,D,E G

No

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Flow Line (Typ'3) g Cnilled Water 4

A,B,E No System Supply Chilled Water System Returu 4

A, B, E No I

Containment Purge Supply 4

A,B,F Yes r

f Hydrogen Purge 4

A,B,D No l

Exhaust

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.j Containment Vacuum 18 A,B,F No "l,

Relief A and B Fuel Transfer Tube 32 A,B,E No Demineralized Water 4

A, B, E No I.

Supply to Contain-ment

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Discharge from Fuel 6

A,B,E No Pool Cooling and Cleanup to Contain-E uant. Pool

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Inlet to Fuel Fool 10 A,B,E No

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Cooling and Clean-up from Contain-

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ment Pool Condensate Makeup 2

A,B.E No j-Supply Dryvell Floor Drain 3

A, B, E '

No Discharge Header L

Containme.nt Ficar 3

A,B,E No Drain Discharge r

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EXHIBIT A

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Line Bypass Considered c

Size Leakage Potential System Service (in.)

Barriers

  • Bypass Path I

Containment Ventilation 36 A,B,F No Air Supply and Exhaust Drywell containment 3

A, B, E No l

Equipment Drains I

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i O Possible Bypass Leakage Barrier Designation :

f I

Isolation valve outside contaica nt A.

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C.

Closed Category I piping system inside containment B.

Isolation valve inside containment D.

Closed Category I piping systen outside containtnant l

E.

Water seal in line

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F.

Line beyond isolation valve outside containment vented to annulus G.

Line teminates outside containment in filtered ECCS area of y

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auxiliary buildin'g L

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l-EXHIBIT B i i' 4 1. 5, l

POTENTIAL UNFILTERED CONTAINMENT BYPASS LEAKAGE PATHS g-Line Descriptio3 Size (in)

Station Air Supply 2

f' Instrument Air Supply 2

l Containment Purge Supply (2) 4 1

Main Steam Line Guard Pipe Feedvater Line Guard Pipe g

Personnel Air Lock I

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EXHIBIT I GUY MARTIN, JR Supervising Engineer Radiological Assessment

SUMMARY

OF EXPERIENCE (Since 1965)

Total Experience - Fifteen years participation i,n Safety Analysis Reports, Environmental Reports, SAR amendments, licensing documents, and cost analysis for insurance premium determination.

Professional Affiliations - American Society of Mechanical Enginers Health Physics. Society American Nuclear Society e

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Intern Engineer in New York State, Certificate No. 022127

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Education -

MS, Polytechnic Institute of New York, 1976

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Nuclear Engineering 3E, City College of the City of New York, School of Harvard University School of Public

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Health, 1977 - Radiological Surveillance Course.

l-REPRESENTATIhE EBASCO PROJECT EXPERIENCE (Since 1973)

l' Supervising Engineer i

Participate in the coordination, technical review and-pre-l-

paration of Safety Analysis Reports (SAR), Environmental Reports (ER), SAR amendments and other licensing documents j

(e.g., Appendix I to 10 CFR 50 studies) for submittal to the Nuclear Regulstory Commission as part of the application for Construction Permit and Operating License of nuclear

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power plants.

_~

Areas of cosplete responsibility include sections of t'i.e SAR dealing with the radiological dose assessment work associated l~

with normal and hypothetical accident conditions.

In this regard, conduct safety reviews-of systems, specifications and operation from a nuclear safety-viewpcint and check their compliance with established nuclear safety criteria.

Furnish techn cal support in the preparation of testimonies for safety hearings and ACES presentation.

Study, develop, i-maintain and use appropriate methods,. including computer programs for evaluating radiological exposures.

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GUY :1ARTIN, JR (Continued)

PRIOR EXPERIENCE (3 years)

Equitable Life Assurance Society of the US Cost Analyst Work involved calculating and analyzing cost of various activities performed throughout the company; assisting departmental managers in their budget preparation work.

Made statistical studies for determination of activity informa-costs and providing company's actuaries support tion for premium determination.

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Dividend Specialist i

Re>'ewed and analyzed lividend and claim reserve cal-cularians.

Prepared disbursement authorizations and Parti-dividend infornation reports for policy holders.

r cipated in training programs for new employees,

?ublications_

r-l-

Martin, G and J Thomas 1978.

Meeting the dose requirements of 10 CFR 100 for site suitability and general design criteria 19 for control room habitability:

a parametric r

l Transactions of American Nuclear Society 24th approach.

Annual Meeting, Vol. 28.

Martin, G, D Michlewice and f Thomas 1973.

Fission 2120:

a program for assessing the need for engineered safety accident feature grade air cleaning systems in post Proceedings of 15th DOE Nuclear Air Cleaning environment.

Conference.

Leti :.a A P, G Martin and J F sip.cy 1979 '. - Implications f

for nuclear facilities of changes heing initiated in the NRC standard atmospheric diffusion model.

Proceeding of rke 41st Annual Meeting of the American Power Conference.

~

R K, Mauro, J, Martin, G.

Effects of Containment

3hacia, Purge on the Consequences of a Loss-of-Coolant Accident.

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Transactions of American Nuclear Society 1980 Annual Meeting L

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Born Philadelphia, Pennsylvania Education Polytechnic Institute of Technology, degree of Engineer in Nuclear Engineering - 1978 Passachusetts Institute of Technology,15 in Nuclear Engineering - 1970 U.S. Coast Guard Academy, BS - 1968 Member American Nuclear Society Licensed Registered Professional Engineer in the State of New York (No. 56673) i

" perience:

1980 Ebasco Services Incorporated, Lyndhurst (NJ) Office; Supervising Engineer, Mechanical-Nuclear Engineering Depar t=ent:

Houston Li 5 ting & Power Co - Allens Creek NGS - Unit No.1 -

3 1200 Mg(e) BWR Technical and administrative responsibility for mechanical, fire. protection, plumbing, HVAC, stress analysis, hangers and Includes supports, and inservice inspection activities.

schedules, budgets, and client relations.

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1978-1980 Ebasco Services Incorporated, Lyndhurst (NJ) Office; Princ# pal Engineer, Mechanical-Nuclear Engineering Department Houston Lighting & Power Co - Allens Cicek NGS - Unit No.1

~~

1200 LW(c) BWR, Lead NSSS Engineer Responsible for preparation and maiatenance of ECCS and BOP flow diagrams, piping layouts, system design descr1ptions, inservice taspection provisions, Ndelear Island building general arrangements, PSAR and FSAR preparation, equipment sizing and specification, NSSS vendor interface for corre-

~'

spondence, drawing review, and contract administration.

Ebasco Services Incorporated, New York Office; Senior Engineer, L_

1976-1978 Mechanical-Nuclear Engineering Department including:

Ie E'ouston Lighting & Power Co - Allens Creek NGS - Unit No.1 -

I200 MW(e) BWR, Ltad NSSS Engineer Louisiana rower & Light Co - Waterford SES Unit No. 3 -

R 1165 MW(e) PWR.

Lead NSSS Engineer 3l (Same responsibilities as listed for 1978-1980 above.)

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1976-1978 R:sponsible for preparation end maintencoce of ECCS cnd (Cont'd)

P.Op flow diagrams, piping layouts, system design descrip-cions, inservice inspection provisions,' Nuclear Island building general arrangements, PSAR and FSAR preparation, equipment sizing and specification, NSSS vendor interface for correspondence, drawing review, and contract adminis-tration.

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1974-1976 United States Coast Guard, Marine Inspection Office, New York; Lieutenant - Supervisory Boil,er Inspector.

Responsibility for supervision, assignment and training of Marine Inspectors in largest Marine Inspection Office

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l in country.

Inspection of hull and machinery material condition of U.S. flag and foreign merchant vessels, and p essure vessels under construction. Application of p

various laws and regulations vf the United States, ASME j

Code, ANSI, TEMA, NEC and NFPA Standards.

Review of engineering plans and alterations, reports from field and resident. inspectors.

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1973-1974 United States Coast Guard, USCGC Spencer (WHEC-36),

' [L Lieutenant - Chief Engineer.

Responsibility for operation, maintenance and repair of hull and engineering plant of 6200 slip twinscrew stea= ship.

Direct supervision of 40

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officers and men.

Duties included preparation of repair specifications and maintenance of vessel records.

Received Coast Guard Achievement Medal for superior performance of duty.

United States Coast Guard, !!arine Inspe'etion Of fice, 1970-1973 New York, Lt and Ltjg - Marine Inspector.

Inspection of hull and machinery of U.S. and foreign flag merchant vessels.

1968-1969 United States Coast Guard, USCGC Mellon (WHEC-717), Ensign, l._

Assistant Engineer Officer.

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Attachm:nt GM/WFM-2 ERRATA Section II, p. 2, delete the second sentence in the first paragraph.

Add the following in its place:

The maximum allowable leakage rate of 0.5 weight percent per day at the calculated peak internal pressure related to the design basis accident is as specified for pre-operational tests in the Technical Specifications.

Section II Containment Leakage Expected for Allens Creek, page 2:

Delete the second paragraph in its entirety and the third paragraph up to "The types of bypass leakage barriers....."

Substitute the following:

"To determine the type of Icakage which may be expected, all containment penetrations are initially considered."

A; Mechanical Penetrations are those penetrations through which piping or tubing enters or leaves che containment.

The penetration assemblies themselves are not consi&ared ac potential bypass leakage paths since they are of welded construction.

Potential leakage through the pipe itself cas, considered.

A listing of piping penetrations is included in updated Table 6.2-12 of PSAR (Amendment No.

59 dated June 1981).

Potential unfiltered leakage paths are also indicated on this table.

Potential unfiltered bypass leakage paths through piping were arrived at by considering the types of bypass leakage barriers for the pipe.

- Pg. 3 add:

" (h) Main Steam Isolation Valve Leakage Control Systems" after item g.

Change the first sentence to read

" Leakage Barriers of types (c) through (h)....."

- Pg. 4 chance "(c) through (g)" to (c) through (h)".

- Pg. 4 add at the end of Section II the following:

" Instrument tubing, other than the list in table 6.2-12, which penntrates the containment are designed consider-ing the guidelines of MRC Reg. Guide 1.11.

Instrument tubes, other than those indicated otherwise in table 6.2-12, are not considered bypass leakage paths since they have a Type "c" or Type "d" barrier".

i on page 4, at the conclusion of Section II, add the following:

B)

Electrical Penetrations are not considered bypass leakage paths since any leakage would he into the Shield Building Annulus.

This annulus is served by the Standby Gas Treatment System.

C)

The Personnel Air Lock and Equipment Hatch will be considered as potential unfiltered leakage pathways and will be tested to 10 CFR S 50 Appendix J Type B criteria.

- Section IV, p. 6 Type C:

Delete the second sentence to the end of the section and add the following at that point:

"The containment isolation valves are indicated in Table 6.2-12 of the PSAR (Amendment 59 dated June 1981).

l All containment isolation valves which have Type "e" bypass leakage barriers will be leak tested in accordance with ASME - B&PV Code Section XI, sub-section IWV, category A requi:ements for leak tightness".

_ Delete Exhibits A,&B and replace with Table 6.2-12 of the Allens Creek PSAR.

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