ML20209H914

Forwards Descriptions of How Alternate post-accident Sampling Sys Will Meet 11 NUREG-0737 Criteria,Discussion of Items Addressed in Util ,Graphs of Analysis Accuracies & One Oversize P&ID

ML20209H914
Person / Time
Site:	Calvert Cliffs
Issue date:	10/31/1985
From:	Lundvall A BALTIMORE GAS & ELECTRIC CO.
To:	Thadani A Office of Nuclear Reactor Regulation
Shared Package
ML20209H921	List: ML20209H914
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM NUDOCS 8511110290
Download: ML20209H914 (23)

Text

.

BALTI MO R E GAS AND ELECTRIC CHARLES CENTER P.O. BOX 1475 BALTIMORE, MARYLAND 21203 ARTHUR E. LUNDVALL. JR.

vice PRESWENT S-October 31,1985 U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D. C. 20535 ATTENTION:

Mr. A. C. Thadani, Project Director PWR Project Directorate #8 Division of PWR Licensing B

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 NUREG-0737 Item II.B.3, Post Accident Sampling

REFERENCES:

(a)

Memorandum from D.

H.

Jaffe, to E.

3.

Butcher, dated August 26,1985, same subject (b)

Letter from A. E. Lundvall, Jr.,

to E. 3. Butcher, dated September 9,1985, Request for License Amendment (c)

Letter from 3.

R.

Miller, to A.

E.

Lundvall, J r.,

dated February 12,1985, Post Accident Sampling at Calvert Cliff.

Gentlemen:

This letter is being forwarded as follow-up to the meeting discussed in reference (a).

Reference (b) forwarded a request for License Amendment which would relocate the Technical Specifications for the Post Accident Sampling System to Section 6.0, Administrative Controls. Reference (c) contained the latest NRC Safety Evaluation for the Post Accident Sampling System at Calvert Cliffs.

The Baltimore Gas and Electric Company indicated in the earlier meeting with the NRC that a different approach to meeting Post Accident Sampling System NUREG-0737 requirements would be proposed by a submittal at a later date. The new approach would simply consist of a capability to obtain and analyze: (1) reactor coolant samples via undiluted manual grab sampling (with dilution capability during analysis) using the NSSS sampling system, and (2) containment atmosphere samples.

No backup method is proposed in either case, since it is our understanding that when no in-line instrumentation is used, a backup method is not required. This submittal represents the technical basis for that approach, and addresses each of the eleven criteria previously addressed in reference (c). Enclosure (1) provides the detailed description of how this approach meets eaca of the eleven criteria.

Enclosure (2) discusses additional items which were addressed in reference (c) and how BG&E proposes to satisfy them.

Enclosure (3) contains graphs of analysis accuracies. Enclosure (4) is a simplified drawing outlining the major components of the proposed PASS.

8511110290 851031 OM k

l Lg DR ADOCK 05 7

Mr. A. C. Thadani

^"

. October 31,1985

' Page 2 l'"

L In summary, we are requesting the NRC approve and issue a revised Safety Evaluation, and the accompanying Technical Specification changes requested in reference (b).

Should you wish to discuss this matter further, please feel free to contact us.

Very truly yours, l q

W

-[

h A. E. Lundvall, Jr.

Ice President - Supply i

l l

AEL/LES/ dim Enclosures cc:

_ D. A. Brune, Esquire G. F. Trowbridge, Esquire D. H. Jaffe, NRC T. Foley, NRC T. Magette, DNR

+

WJ i-l-i r

p Enclosure (1)

NUREG-0737 CRITERIA

' CRITERIA

RESPONSE

-(1)

The-licensee shall have the capability to (1) a.

The time required for sampling and promptly obtain reactor coolant samples analysis of reactor coolant and and containment atmosphere samples.

containment atmosphere samples

-The combined time allotted for sampling will be:

and analysis should be three hours or less from the time a decision is made to take RCS PASS:

a sample.

Technician briefing / dress 40 min.

PASS sampling 15 min.

Purge Time

• 45 min.

Sample Preparation 10 min.

Analysis 30 min.

TOTAL:

140 min.

• This assumes a flow of 900 cc/ min. The purge time is dependent on the sample flow rate.

Containment Atmosphere:

Technician briefing / dress 40 min.

Preparation 10 min.

Purge time 15 min.

Analyses 20 min.

TOTAL:

85 min.

b.

A loss-of-offsite power is not a design basis event for which post-accident sampling must remain operable.

The implications of not having the capability to draw a sample during a loss-of-offsite power event have been considered.

The reasons why the post-accident sampling system was not designed

~

with the capability to function during a loss-of-offsite power event are as follows:

1.

The post-accident sampling function is not safety-related.

The information provided by analyzing reacter coolant samples after an accident involving the potential for core

NUREG-0737 Pags 2

~

CRITERIA:

RESPONSE

t damage -

serves' only to,

complement and confirm other sources of 'information which'-

allow a. determination. to - be made in the Technical Support -

. Center and -the Emergency Operations Facility. regarding the ' degree of core damage. The -

other sources of information-

' available include the i contain- :

ment -. high' range. radiation.

monitor, 'which - as the greatest.

value, and the containment '

hydrogen analyzers,- which aret important to a somewhat lesser

. degree.'.'Both of these sources ~of.

information regarding the extent -

of ' core ' damage L are. safety-related; - they : also will-be available-early in an. accident scenario, including one which -

involves a

' loss-of-offsite power. -

~Thus information-obtained via coolant ; sample analysis ~ is only confirmatory with regard to the extent of core -

damage and is not required to -

support decisions - during 'the early stages

'of accident mitigation or recovery. This is evident ; in

'some of - the s

parameters measured through coolant analysis:

Radioisotopic analysis:

This provides - specific-.. information

- relevant to the physical state of the fuel; i.e.,' the degree to which the fuel was overheated and subjected to metallurgical

. transformations which

_ i

. facilitated the release of various isotopies into the coolant.' This information is relevant to

~

- decisions made regarding ' long term -

cooling modes -

and reactivity control, ' but is ~ not E

relevant. to' any immediate or near term operator actions.

Boron Concentration:

This is relevant only to assure adequate.

shutdown margin.

wt y+

4 e+

v.-

,,e s.~+

--%m.

~.y-,, <

-- ~,.

~

1NUREG-0737 '

Pag 2 3l 0

CRITERIA ~

RESPONSE

~

Chloride Concentration:

Relevant to long term RCS integrity.

Hydrogen Concentration: Long-term RCS integrity, extent of fuel-cladding damage (supplements H2 analyzers).

In short, a lapse _ in. coolant analysis-derived information during a loss-of-offsite power

, event is not inimical to public health and safety since this is confirmatory information, which only supports long-term recovery activities and decisions, and it is expected that offsite power sources will be ' available -. Very -

shortly after initial loss.

2.

Loss-of-offsite power events are infrequent and are historically very short 'in duration.

The~

current frequency of loss-of-offsite power events nationally is about.045 events / site-year or roughly one event per-20 years.

On the average, offsite power is testored 'within - the first 30 minutes and ninety (90) percent

-of the events are terminated within three hours.

In comparison, it is. typically not expected that a decision would be made to take an RCS sample (and _ Rad. _ Safety-surveys completed to support- -the activity) until one to two hours af ter a small-break LOCA at the very earliest. The NRC Staff's guidance then allows three hours to accomplish the activity.

However, as discussed above, coolant sample analysis :is not likely to provide _ information relevant to recovery functions until many hours or days following the event.

This was confirmed by the experience at TMI-2.-

NUREG-0737 -

Paga4 CRITERIA

RESPONSE

For the above reasons, it is our position - that additional design provisions to make the PASS operable under loss-of-offsite power conditions are not necessary.

(2):

The licensee shall: establish an onsite (2) a.

An aliquot of the diluted sample

- radiological _. ~ and chemical analysis shall be diluted further if necessary, capability -to provide, within three-hour and a gamma spectrum shall be time.

frame ~. established above,.

conducted on this by use of the quantification of the following:

chemical laboratory's gamma analysis equipment.

The a.

Certain radionuclides in the reactor radionuclide

- analysis of the coolant and containment atmosphere containment atmosphere shall be

that may be indicators of the degree conducted by obtaining a syringe of ~ core damage (e.g., noble gases; sample of the containment

'lodines and cesiums, and non-volatile atmosphere and analyzing an aliquot isotopes);-

of this mixture on the chemical laboratory's gamma analysis b.

Hydrogen levels in the containment equipment.

atmosphere; b.

Hydrogen levels in the containment c.

Dissolved gases (e.g., H ), chloride atmosphere - shall be determined by 2

(time allotted for analysis subject to the use of an on-line hydrogen.

discussion '

below),

.and boron analyzer.

concentration of liquids.

c.

Dissolved gases can be obtained from d.

Alternatively,.

have in-line the RCS PASS.. The dissolved gases

' monitoring capabilities to perform are stripped from the liquid and a

all or part of the above analyses, sample of this-can be obtained by means of a syringe.. This sample can then be injected into ~ the ~ gas partitioner for hydrogen analysis.

The~ concentration of these gases in the-liquid can then be obtained via established calculational methods.

The chloride analysis ~can be obtained by means of a

Dionex ion chromatograph.

Since 4 minimum.

sensitivity of. this is 2 - 5 ppb, an analysis of this can be done with a minimum dilution of 30:1 to meet 150 ppb criteria. Since the minimum boron shutdown margin is 350 ppm, a 30:1 dilution of the sample can also be performed.

Established procedures can~ meet this sensitivity.

~

l 7

NUREG-0737 Paga 5 -

CRITERIA

RESPONSE

'(3)

Reactor coolant and containment (3)

The PASS valves which are not atmosphere. sampling during post accessible after an accident are designed accident conditions shall not require an to remain operable under the isolated. auxiliary system (e.g.,

the environmental conditions in which they letdown ' system, ~ reactor water cleanup need to operate.

system ' (RWCUS) to - - be placed in operation in order - to use sampling The sampling of the reactor coolant, low system.

pressure safety injection - or the containment atmosphere does not require any isolated auxiliary system to be put in operation. All remotely operated valves which must be operated are. powered from safety-related power sources and receive air from the safety-related saltwater air compressors.

(4)-

Pressurized reactor coolant samples are (4)

This Criterion is satisfied by-the not required if the licensee can quantify explanation given in 2(C).

the amount of dissolved gases with unpressurized - reactor coolant samples.

The measurement of either ' total

' dissolved gases or H2 gas in reactor coolant samples is considered adequate.

Measuring the 0 concentration is 2

recommended, but is not mandatory.

(5)

' The time for a chloride analysis to be (5)

Although we.do not have to analyze for performed is - dependent upon two chloride for ninety-six hours, the analysis factors: -(a) If the plant's coolant water can be performed within the initial is ' seawater or brackish water and (b) if sampling time frame stated in Criterion there is' only a single barrier between (1).

primary containment systems and the cooling water. Under both of the above conditions 'the licensee shall provide for a chloride analysis within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the sample being taken. For all other cases, the licensee shall provide for the analysis to be completed within four days. The chloride analysis does not have to be done onsite.

(6)

The design basis for plant equipment for (6)

The radiation exposures to an individual reactor coolant and containment for the sampling and analysis of the atmosphere sampling and analysis must reactor coolant and-containment assume that it is possible to obtain and atmosphere samples will be:

analyze a ' sample L.without radiation exposures to any individual exceeding the criteria of GDC 19 (Appendix A,10 CFR L

m NUREG-0737 -

Pagri61

~

P CRITERIA

RESPONSE

.m

. Part 50) (i.e., _5 rem whole body, 75 rem RCS PASS

> extremities). - (Note that the design and operation review criteriont was changed Whole Body 0,63 rem t

.from the operational limits' of '10 CFR Extremities 8.55 rem Part 20 (NUREG-0578) -to the'GDC -19 criterion (October. 30,1979 letter from-H. R. Denton to all' licensee,s))..

CONTAINMENT ATMOSPHERE Whole Body 0.5 rem

^

Extremities.

1.6 rem

= The above exposures assume no dilution during sample collection and a 5:1

' dilution during sample analysis.

- (7).

The analysis ofprimary coolant samples (7)

- The - analysis will be performed as

~

for _ boron.is required for _PWRs. (Note outlined in 2(c).

, that Rev. 2 lof Regulatory Guide 1.97

. specifies the need for primary coolant boron analysis capability at BWR plants).

(8)

If - in-line t monitoring is used for any (8)

In-line monitoring will only be used for sampling - 'and analytical capability the containment hydrogen analysis -

.specified herein, the licensee shall consistent with NRC

" guidance" provide' backup sampling through grab Technical Specifications provided in samples,' and shall demonstrat_e the Generic Letter 83-37.

capability of analyzing -the samples.

-Established planning for._' t analysis at offsite -

facilities-is -

acceptable.

Equipment provided for backup sampling shall be capable of providing at least one sample per day for seven days following onset of the accident, and at least one sample per. week until' the' accident condition no longer exists.

.(9)J

- The' licensee's radiological and chemical (9) a.

Liquid: Although the initial sample.

sample analysis ~ capability shall include will be 5:1 dilution, a 0.1 mi from provisions to:

the sample syringe can then be injected to a one liter volumetric a.

Identify and quantify the isotopes of

flask, obtaining '

an additional

the < nuclide. categories discussed 10,000

1 dilution. Additional similar above the levels corresponding to the dilutions can be performed.

source terms given in Regulatory

' Guide 1.3 or 1.4 and 1.7. Where r

e 3,

ENUREG-0737~

l Pago 7_-

CRITERIA'

RESPONSE

~

necessary and practicable,. the' Containment Atmosphere: -The grab.

ability to dilute samples to_ provide sample from -- the containment

~

capability = for measurement and atmosphere can be diluted by reduction. of _ personnel - exposures injecting a 0.1 -mi sample into an should be provided.

Sensitivity' of =

1four liter marinelli. providing a onsite

' liquid

. sample analysis 40,000:1 dilution. If further dilutions capability; for measurement. and are,needed,: additional similar reduction _ -of.. personnel, exposure dilutions ~ can be performed.

should be provided. = Sensitivity of onsite liquid sample analysis b.

For low - levels - of -background capability should.be such as to interferences, a background subtract.

permit. measurement of radionuclide routi_ne can - be performed.

If the concentration. In ' the. range ~ from radiation background levels. are approximately lu Ci/g to 10 C1/g.

excessive the samples can. be counted in our mobile laboratory.

b. - Restrict background-levels of which has complete gamma counting radiation In the radiological and facilities.

chemical analysis facility from sources such that the sample analysis

will provide results with ' an acceptably.

small error (approximately a factor of 2). This can be accomplished through the use of. sufficient., shielding around

~

~ samples and outside sources, and by the use - of _ a ventil tion system a

design : which will control ~ the-

. presence of airborne radioactivity.

The gamma analysis is done utilizing

(10)- - Accuracy, range and sensitivity shall be.

(10)

a..

! adequate to provide pertinent data to the the

_ Chemistry Laboratory operator _ in order to describe radiological multichannel' analyzer which is

-and chemical. status of the reactor calibrated with NBS-traceable

- coolant systems, standards. This analyzer is routinely checked

'by.

the.NRC Mobile

a. _ L Gamma ~ Spectrum - Clarification:

Laboratory with spiked or split Gross _. activity,. gamma spectrum:

samples. In addition, the analyzer is measured. to estimate core damage, checked weekly against standards -

these analyses should be accurate using the NRC "R" test or similar Lwithin a factor of two across the criteria.

A copy.of the "R" test entire range. -

criteria - is attached for.your information.

b. - Boron: _ In-general this analysis should be accurate within15% of.the measured value. For concentrations

- below'1000 ppm the tolerance band should remain at150 ppm.

4 4

L NUREG-0737J

' Pag? 8 CRITERIA

RESPONSE

c. - Chloride -

Clarification:

For b.

The lowest boron concentration to.

concentrations between - 0.5 and obtain required shutdown margin 4

20.0 ppm chloride the analysis should under.the worst conditions is

. be ~ ' accurate i within 110% of the 350 ppm ' boron. _A. determination of

measured values. At concentrations the. accuracy and linearity of the

.below 0.5 ppm the-tolerance band boron concentration was determined

' remains at1 05 ppm.

in the following test matrix:

0 d.

Hydrogen.. Clarification:

' An 40 ppm Potassium Iodide

-accuracy of 110 % is desirable 250 ppm Cesium Nitrate between 30 and 2000 cc/kg but120 %

10 pp.m Barium Nitrate can-

'be.

. acceptable.

.For 5 ppm Lanthanum Chloride concentrations below 50 cc/kg the 5 ppm Ammonium Cerium Nitrate tolerance remains at1 0 cc/kg.

2000 ppm Boric Acid 5

2 ppm Lithium Hydroxide

e. -pH Clarification:. Between a pH.of 5 1100 ppm Trisodium Phosph.

to 9, the reading'should be accurate..

12 H O 2

within 1 3 pH units. For all other 0

. ranges 1 5 pH units is acceptable.

The attached graph shows the 0

accuracy of our analyses in this test matrix using a 30:1 dilution as compared to NUREG 0737 recommended ranges. For the lower limit of detection our diluted sample showed a 10 1 1.3 ppm deviation which corresponds to 300 1 39 ppm boron in the sample.

I c.

The lowest chloride concentration of concern is 0.15 ppm.

A determination of the accuracy and linearity of the chloride concentrations was determined in the following test matrix:

40 ppm Potassium Iodide 250 ppm Cesium Nitrate 10 ppm Barium Nitrate 5 ppm Lanthanum Nitrate 5 ppm Ammonium Ceruim Nitrate 2000 ppm Boric Acid 2 ppm Lithium Hydroxide 1100 ppm Trisodium Phosphate

• H O 2

The attached graphs show the-accuracy of our analysis in this test matrix using a 30:1 dilution as compared to NUREG 0737 recommended ranges.

The interference of the Trisodium

NUREG-0737

Pag 2 9 c

.o - <

CRITERIA

RESPONSE

" Phosphate -12 H 0 caused' four-7 separate linear regions as indicated by the graphs. For the lower limit of ~

c detection of our diluted sample ~

showed' a 0.005 1 0.0003 ppm

' deviation which corresponds to a 0.15: 1 0.009 ppm chloride in the sample.

+

d.

The lowest hydrogen concentration of concern is 10 cc/kg.

A determination of the accuracy and linearity of the hydrogen concentration was. made.

The attached graphs show the accuracy of the analysis using a 50cc. dilution volume in the PASS as compared to NUREG 0737's recommended -

-- ranges.

For the lower limit of detection our diluted sample showed a-10 1

.0.4 cc/kg hydrogen concentration ' ' in the Reactor Coolant System.

m e.

A -laboratory analysis-of the pH of various concentrations of boric acid -

and trisodium phosphate

• 12 H O -

2 were determined in the standard test -

martrix of:

40 ppm Potassium Iodide 250 ppm Cesium Nitrate 10 ppm Barium Nitrate 5 ppm Lanthanum Chloride 2 ppm Lithium Hydroxide Since the trisodium phosphate 12 H O is a chemical additive which 2

is presently ' covered by Technical Specifications, credit for this should be permitted. Although our FSAR takes credit for 925 ppm Trisodium

- Phosphate

• 12 H 0, the attached 2

graphs show that the pH does not go below 6.5. Concern over the coolant corrosion potential is not significant-until acidic conditions exist.

Therefore, the concern over - the measurement of pH is not justified and it is our position that the

'? NOREG-0737

- lY Paga 10 s

?

CRITERIA

RESPONSE

^

requirement' to analyze the sample for pH should be eliminated for our plant.' '

(11)) jInithe ; design; of the post accident (11)

Since the' NSSS ' sampling system 'will be sampling _. -and~. analysis capability,-

utilized as the PASS, manyy of these

consideration should be. ~given to the considerations have- ' already been.

. following items:

submitted, reviewed and approved by the NRC as meeting the' = interim -- PASS L a.'

Provisions for purging sample lines, requirements.

for reducing plateout in sample lines, 1-a.: We have demonstrated that this Post-ifor -Jminimizing sample. ; loss = 'or

-Accident Sampling method does not.

distortion, ' for-. preventing blockage, foi sample lines by-loose material in have a sample plateout problem by

. the RCS ~

containment, : 'for comparing samples. drawn from or-

. anpropriate disposal 'of the sarnples,~-

different locations and. through and ;for flow restrictions - to limit-different sampling systems..

. reactor coolant = loss from a rupture

~ /2" tubing of,the sample.line.

The post The sample lines are 1

accident

. reactor coolant. and which. allows for a relatively high

. containment 1 - atmosphere samples sample velocity to assist in reducing

~

should be representative of~ the plateout as well as limit the volume

~

. reactor coolant in the core area and

. of reactor coolant loss from a the containment -

atmosphere rupture of the sample ~ line.

The following a transient ' or accident.

reactor coolant sample 1 is drawn

~

The sample lines should be as short from : the normal reactor coolant as possible.to minimize the volume sample line using the RCS pressure of -

fluid to be. taken from as a driving head. In the event of

containment. The residues of sample insufficient pressure in the RCS, the collection should be' returned to sample ' will. be - drawn from the.

containment or to a closed system.

discharge header of;the LPSI pump.

The former assures a representative

' b.

The ventilation : exhaust.from the sample in the. case of a small LOCA E

sampling ~' station should be filtered and the latter

. assures a.

l'

' with charcoal absorbers and high-representative -- sample' via the efficiency particulate' air (HEPA) containment sump in the event of a filters.'

large LOCA.

The reactor coolant

-and LPSI samples will be returned to the affected ' containment ' af ter a 1

modification is completed.

The containment atmosphere sample.

y~

originates from the containment via the normal hydrogen sample'line at the 135' elevation. Except for that which is removed for analysis ~ by syringe, the sample is returned to containment via the normal hydrogen L

sample line.

The containment isolation valves associated with the sampling system close automatically on a SIAS signal.

p F

f

'NUREG-0737 Peg 211 CRITERIA'

RESPONSE

b.

The. degassing station in the chemistry laboratory will be equipped with charcoal filters. The laboratory ventilation system is processed through HEPA filters in the Auxiliary Building and Waste Processing Ventilation system.

ENCLOSURE (2) o I

ADDITIONAL ITEMS

.(1) bore Damage Assessment Procedures Core Damage Assessment Procedures submitted in our letter dated December 31, 1984, will - continue to be utilized.

A resubmittal of these procedures is

' unnecessary.

(2)

Training The technicians designated to operate the PASS are trained and qualified.

Training is normally scheduled for all candidates and qualified technicians approximately every six months. Training includes:

~(1)

A functional description and layout of the system and components, (2)

Operations, procedures, valve positions and flow paths, and l-(3)

Limitations, setpoints, protection devices, and dose rates.

I

ENCIMURE (3)

RCP-2-101 ATTACHMENT (2) g;h WEEKLY EFFICIENCY CHECK R-TEST

~

DATE' ND#

SCAN # _

FOR NBS SOURCE #

PTS. EVALUATED Cd-109; Cs-137; Co-60 (1.173 Mev peak) 50 TOPE ACTIVITY ERROR Cd-109

+ /-

Cs-137

+ /-

Co-60

+ /-

R-value =

Activity

+ 0.5

% limit =

Activity 2 o error Source Activity Cd-109 R value

=

=

+

=

% Limit =

=

Cs-137 R value

=

+

=

% Limit

=

Co-60 R value

=

+

=

m

% Limit

=

Ratio Ratio Ratio Possible Possible Agreement Agreement A Agreement B

<3 0.4 2.5 0.3 3.0 No Comparison 4-7 0.5 2.0 0.4 2.5 0.3 3.0 3 - 15 0.6 1.66 0.5 2.0 0.4 2.5 16 - 50 0.75 1.33 0.6 1.66 0.5 2.0 51 - 200 0.80 1.25 0.75 1.33 0.6 1.66

> 200 0.85 -

1.18 0.80 1.25 0.75 1.33 "A" criteria are applied to the following analyses:

• Gamma Spectrometry where principal gamma energy used for identification is greater than 250 Kev.

• Tritium analyses of liquid samples.

"B" criteria are appiled to the following analyses:

• Gamma Spectrometry where principal gamma energy used for identification is less than 250 Key.

UNSATISFACTORY (not in ratio agreement)

Technicians Signature gg Technicians Signature PRINCIPLE TECHNICIAN REVIEW Signature Rev.6

BORON (STANDARD)' ppm VS.

BORON (CALCULATED) ppm FROM MANNITOL TITRATION ALLOWABLE EXPERIMENTAL-THEORETICAL TOLERANCE DEVIATION Boron Calculated (ppm) 100

// /

,.,s /,..

/ r: -

. / e,.

,. s //

.s s 80

//. // *

,. /,sf,s,.

. //.. *

.s s f,,,sf

• /<,.

/'.

80

. s,s.-

. s s.'

//

• 0's.'//

. /..,*N

/ s, 40

/

l ~/

20

,. ~.

~

f 0

O 30 80 B0 120 150 Boron Standerd (ppm)

(Data is based on a 30: 1 dilution of reactor coolant)

+

CHLORIDE ANALYSIS BY HPIC IN P.A.S.S.

MATRIX (1: 30 DILUTION)

ALLOWABLE THEORETICAL EXPERIMENTAL TOLERANCE DEVIATION Peak Height (cm)

I a

l e

7-

~

w

,,,~~~~~~~~~~x 8

~

~~

~~~~~~~~~

6 e

~

__ _____~~~~~~~~

i i

4 l

}

___ ___~~~~~~~~~

s

~

r

~~

~~~~~

5

~

............. ~. __~~~~~ ~~,,,,,,,,,,,,,,

i

~~~~~~~~..... ~~

2

~

7 y

/

=

r.....,,,,,,,,,

r r

1 1

r 1

i t

r i

r t

0 O

4 6

e 7

8 a

to 22 2g ta 24 26 C1 Concentration (ppb) i Note: All peak heights are normalized to a 0.3 micros output range i

CHLORIDE ANALYSIS BY HPIC IN P.A.S.S.

MATRIX (1: 30 DILUTION)

ALLOWABLE THEORETICAL EXPERIMENTAL TOLERANCE DEVIATION

$O4; Jg

'T 4e

~

g 4.

A n........

pg

~

L

""..u Jg

~

~

14

...u..*"....,,,,,,,,,,

~

1. 4

~

~

10 g

~

o

." " n g

5' -

4 3

2 t

0

--20 30 40 50 0

10 i

j C1 Concentration (ppb)

Note: All peak heights are normalized to a 0.3 micros output range

CHLORIDE ANALYSIS BY HPIC IN P.A.S.S.

MATRIX (i: 30 DILUTION)

ALLOWABLE THEORETICAL EXPERIMENTAL TOLERANCE DEVIATION Peak Height (cm)

~

... -~

........~-

ts

~

.......~~~~

20 s

e

......... -~-,,,,,,,,,,,,,,,

~.

e w

• = = =..

.: = = =

~

g.............

~

e W

5. -

t l

I I

1 I

4 100 40 50 80 70 80 90 C1 Concentration (ppb) -

Note: All peak heights are normalized to a i micros output range

CHLORIDE ANALYSIS BY HPIC IN P.A.S.S. MATRIX (1: 30 DILUTION)

ALLOWABLE TEGETICAL EXPERIENTAL TOLERANCE DEVIATION Peak Height (om) s

~

s

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

$' u................"....

=

a s

a **n s

rr

~

$a J;;";;;.............

s

"ll;..............

v s

E "...... ".........

~

tK :u...S...

S..

4

~

meaxh s

1 4

s s

t f

a f

O k

k k

%lp

/

C1 Concentration (ppb)

Note: All peak heights are normalized to a i micros output range s

.s HYDROGEN (STANDARD) cc/kg VS.

i HYDROGEN (CALCULATED) cc/kg FROM GAS CHROMATOGRAPHY i

EXPERIMENTAL THEORETICAL ALLOWABLE DEVIATION TOLERANCE 4

1 r

~

Hydrogen Coloulated oc/kg 50 l

/ /

-/

/ /

/ /

f

,# s/ /

s l

,s

/

/ / '/ /

/ y.i

./ /

40

/ /

y / /...'

y s s

./

/,s

.f s s p

,s f

l f

//

30

/

//

/

p

/s/.

s s s

/

//

/

/

/ s/ /

s s/

/

,'s' l

20 f

/

/

</

- y< y.

f e

l

//

l

+,+

/

10 y

f

.J i

f J

a e

a a

a a

a e

a a

a O

25 50 75

)

Hydrogen Standard co/kg l

(Data is based on a dilution with 15 liters of Nitrogen)

I t

k HYDROGEN (STANDARD) cc/kg VS.

HYDROGEN (CALCULATED) cc/kg FROM GAS CHROMATOGRAPHY EXPERIMENTAL THEORETICAL ALLOWABLE DEVIATION TOLERANCE Hydrogen Calculated oc/kg s

/

/s/

s

,.! !/,f

-, //

/ // /

/ / *.*

1500

//

,. // y/

• //./

s/s

,. ' /$?

l J /. *,.

, py..

i

. #s.

,/ s?y'/

1000 l

<p

,',/

j j

/

500 t

0 O

500 1000 1500 2000 2500 3000 i

l Hydrogen Standard oc/kg (Data is based on a dilution with 15 lite.'s of Nitrogen) i

pH FOR BORIC ACID AND TSP SOLUTIONS IN TEST MATRIX 0 ppm 825 ppm 1250 ppm 2500 ppm Ne3PO4 x 12H2O No3PO4 x 12H2O Ne3PO4 x 12H2O Ne3PO4 x 12Hx0 i

=

0

)H 12 3

ii 1

10 p

1 8

1 8 r 1

7. r 1

a 8

r 1

5 3000 O

500 1000 1500 2000 2500 BORON

OVERSIZE

' DOCUMENT PAGE PULLED SEE APERTURE CARDS NUMBER OF PAGES:

ACCESSION NUMBER (S):

1 25//lln 29f APENTURE CARD /MARD COPY AVAILABLE FROM RECORD SERVICES BRANCH,TIDC FTS 492-8939 I

i

-,--,-.--,,-----n-,.--__,

______,_...-.-.,,,n,,