ML20113C870

Informs NRC of Changes to Past Commitments Made in Response to NUREG-0737,Item II.B.3,Post-Accident Sampling Capability

ML20113C870
Person / Time
Site:	Grand Gulf
Issue date:	06/27/1996
From:	Hutchinson C ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM GNRO-96-00076, GNRO-96-76, NUDOCS 9607020151
Download: ML20113C870 (22)

Text

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Enttrgy Cpersti:nt,Inc.

' = ENTERGY eo sox 7so Ptyt Gibson.MS 39150 Tel 601437 2800 C. R. Hutchinson Voe Preset Oprators June 27, 1996 "d " **""

U.S. Nucl'ar Regulatory Commission Mail Station Pl-37 Washington, D.C.

20555 Attention:

Document Control Desk

Subject:

Grand Gulf Nuclear Station Docket No. 50-416 License No. NPF-29 Notification of Changes to the Post-Accident Sampling Program GNRO-96/00076 Gentlemen:

The purpose of this letter is to inform the Nuclear Regulatory Commission of changes to past commitments made in response to NUREG-0737, Item II.B 3, Post-Accident Sampling Capability.

Specifically, a, number of commitments were made in response to NUREG-0737, Item II.B.3 that have been reviewed and found to be no longer supportive of meeting thq objectives of NUREG-0737 and Regulatory Guide 1.97.

As a result of these findings, Entergy intends to simplify its post-accident sampling program with the objective of eliminating unnecessary program elements.

Pursuant to 10CFR50.59, Entergy plans to revise Sections 7.7.1.11, 7.5.1.2.18, 9.3.2; Tables 7.5-2 and 9.3-3 of the Updated Final Safety Analysis Report (UFSAR) and the post-accident sampL.ng system (PASS). The following briefly describes the program deletions:

1)The sample points from the jet pump diffuser and the reactor recirculation system; 2) The commitment to take a drywell atmospheric sample using the Post Accident Sampling System to determine hydrogen concentration during accident monditions; 3) The commitment to determine pH, and the concentration ef dissolved gases and chloride in the reactor coolant; 4) The commitmen; to take a drywell atmospheric sample in order to estimate the deqree of fuel cladding failure.

The above items do not mitigate the consequences of an accident; nor do they ensure the operation, manually or automatically, of any safety component or system. These items are not essential to safe operation of the power plant in a post-accident condition.

Due to there being other sample points that can be used to obtain reactor coolant, the two sample points mentioned above are being deleted. Hydrogen concentration can be determined by other qualified accident range instramentation.

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June 27, 1996 GNRO-96/00076 Page 2 of 3 I

Determination of pH would indicate a possible corrosive environment.

i However, this information would not be useful to prevent or mitigate l

the consequences of an accident. Dissolved gas and chloride analyses (which are also used to determine the corrosiveness of the reactor environment) en reactor coolant were required due to reactor vessels l

not having adequate depressurization capability. Subsequent requirements to have vessel venting capabilities, make these analyses no longer necessary.

Core damage estimation can be performed using other qualified instruments, as well as alternative liquid sampling points.

Therefore obtaining an atmospheric sample from the drywell would not be necessary.

l Our evaluation concluded that these changes did not involve any safety concerns. The deletion of the above commitments would not adversely impact the mitigation of an accident, safe shutdown of the reactor or the ability to estimate core damage.

The changes identified in this letter to the PASS program do not in any way decrease the effectiveness of the PASS program in meeting the objective of NUREG-0737, Item II.B.3.

The UFSAR revision is provided as an attachment to this letter.

Please direct any questions pertaining to this effort to Riley Ruffin at (601) 437-2167.

Yours truly, g

a-

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L/

CRH/RR/

attachments:

1.

References 2.

GGNS UFSAR Sections for Revisio.,

cc:

Mr. J.

E. Tedrow (w/a)

Mr. R. B. McGehee (w/a)

Mr. N. S.

Reynolds (w/a)

Mr.

H. L. Thomas (w/o)

Mr. J.

W. Yelverton (w/a)

Mr.

L. J. Callan (w/a)

Regional Administrator U.S.

Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 Mr.

J. N.

Donohew, Project Manager (w/2)

Office of Nuclear Reactor Regulation U.S.

Nuclear Regulatory Commission Mail Stop 13H3 l

Washington, D.C.

20555

, to GNRO-96/00076

References:

1. NUREG-0737, " Clarification of TMI Action Plan Requirements," USNRC, 1

November 1980, Washington, D.C.

2. Letter to H.

R. Denton (NRC) from L.

F. Dale (MP&L), September 1,

1981,

" Grand Gulf Nuclear Station Units 1 and 2, NUREG-0737, Item II.B.2, l

Plant Shielding, NRC Docket Nos. 50-416/417."

i

3. Letter to H.

R. Denton (NRC) from L.

F.

Dale (MP&L), April 15, 1982,

" Grand Gulf Nuclear Station Units 1 and 2, Response to SER Item 1.11(17), NRC Docket Nos. 50-416/417."

4. Letter to H.

R. Denton (NRC) from L.

F.

Da.e(MP&L), October 23, 1981,

" Grand Gulf Nuclear Station Units 1 and 2, Response to SER Items 1.10(22), 1.11(17) and 1.9(8), NRC Docket Nos. 50-416/417."

5. Letter to H.

R. Denton (NRC) from L.

F.

Dale (MP&L), December 18, 1981,

" Grand Gulf Nuclear Station Units 1 and 2, Additional Information to NRC Question 281.9, NRC Docket Nos. 50-416/417."

6. Letter to H. R. Denton (NRC) from L.

F.

Dale (MP&L), July 30, 1982,

" Grand Gulf Nuclear Station Units 1 and 2, Post Accident Sampling, Operating License Condition 2.C. (44), NRC Docket Nos. 50-416/417."

7. Letter from W.

T. Cottle (SERI) to NRC, July 27, 1989, " Grand Gulf Nuclear Stetion Units 1 and 2, PASS Sampling per SSER4, NRC Docket Nos.

50-416."

8.

Letter to O.

D.

Kingsley, Jr. (SERI) from D. M. Collins (NRC), November 24, 1987, " Grand Gulf Nuclear Station Unit 1, NRC Inspection Report No.

50-416/87-30, NRC Docket Nos. 50-416."

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2 7.7.1.11 Process Sampling System - Instrumentation and Controls 7.7.1.11.1 System Identification 7.7.1.11.1.1 General The purpose of the process sampling system instrumentation and controls is to collect representative liquid and gas samples for analysis and to provide analytical information required to monitor plant and equipment performance and changes to operating parameters.

7.7.1.11.1.2 classification This is a power generation system and is classified as not related to safety.

7.7-59 Rev. 0 1.

~ _ _ _. _ _

GG UFSAR 7.7.1.11.2 Power Sources The PSS instrumentation and controls are powered from non-essential buses.

7.7.1.11.3 Equipment Design 7.7.1.11.3.1 General The process sampling system is described in subsection 9.3.2.

7.7.1.11.3.2 Testability Since the process sampling system is usually in service during plant operation, satisfactory performance is demonstrated without the need for special inspection or testing beyond that specified in the manufacturer's instructions.

7.7.1.11.4 operational Considerations 7.7.1.11.4.1 General Information

(

A method for relating radionuclide gaseous and ionic species to estimate core damage has been devised by General Electric for the fuel type being used at Grand Gulf.

This methodology was l

utilized in the development of a core damage procedure which is currently in effect at Grand Gulf.

The process sampling system l

is manually operated at grab sample panels located throughout the plant.

The sampling panels are designed to* minimize contamination and radiation at the sample station.

Appropriate shielding and area radiation monitors will minimize radiation effects.

7.7.1.11.4.2 Post-Accident Sampling Station The post-accident portion of the process sampling system is designed with the following operational considerations:

l a.

The system is capable of obtaining and analyzing reactor coolant and containment atmosphere samples l

within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> from the time a decision is made to take a sample.

l b.

Facilities are provided on' site to perform the analysis i

described in subsection 9.3.2.2.4, within the 3-hour time frame established above, except fer chieride "hich must bc :n:1 ::d within i d:y: (OS heur;).

5 7

f c.

Reactor coolant and containment atmosphere sampling

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during post-accident conditions does not require an l

isolated auxiliary system to be placed in operation in order to use the sampling system.

I 7.7-60 Rev. 5 12/90 e

GG UFSAR d.

Analyses for pHy conductivity, di?5cived hydregen, dire 1 red crygen and containment

="4 dry ell atmosphere oxygen / hydrogen concentrations may be performed by using the installed PASS process instruments.

The other post-accident analyses described in subsection 9.3.2.2.4 including chlorid: will be performed via grab sampling.

e.

The post-accident sampling station is designed to provide adequate radiation protection so that it is possible for an operator to obtain and analyze a sample without radiation exposures exceeding the criteria of GDC 19.

is f.

The system's primary method of sampling will be grab sampling.

The equipment provided for sampling is capable of providing at least one sample per day for 7 days following onset of the accident and at least one sample per week until the accident condition no longer exists.

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g.

Through cy:::r crd laboratory dilutions, the radiological and chemical analysis capability of laboratory equipment includes provisions to identify and i

quantify the isotopes of the nuclide categories of 4

concern to levels corresponding to the source terms given in Regulatory Guides 1.3 and 1.7, i

h.

Sy:::= ned laboratory dilution 4 allow # the counting l

room's gamma detection system the abilfty to monitor j

reactor coolant activity over the ranges required.

]

Liquid geometries and laboratopy germanium detectors I

.re vill be used to meet 10 to 10 C1/cc reactor coolant i

and containment atmosphere of 10' to 10' Ci/cc.

4 i.

The post-accident sampling system instrumentation pro-vides adequate ranges, accuracies, and sensitivities to allow the operator to obtain pertinent data in order to describe the radiological and chemical status of the reactor coolant systems.

In addition, the ranges of the instruments meet those ranges identified in Regulatory Guide 1.97, Revision 2 except as agreed upon by the NRC in letter dated December 6, 1985.

j.

Reactor coolant sample lines'are of a diameter such that the rupture of a sample line will limit reactor coolant loss.

l 7.7-61 Rev. 9 12/95

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UFSAR k.

The post-accident sampling system'(PASS) components l'

required for drawing and analyzing a sample of the reactor coolant and the dr~eell end containment I

atmosphere are powered from a Class 1E motor control center (MCC).

To provide isolation between the Class i

1E power supply and the non-Class 1E post-accident sample system load, the Class 1E MCC will be shed on a

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1 7.7-61a Rev. 5 12/90

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GG UFSAR LOCA signal and will require operator action to be manually reconnected.

The Class 1E MCC will be shed on a LOSP signal; however, i. will be automatically reconnected when the Class 1E power supply is restored.

7.7.1.11.4.3 Operator Information

'The most important process streams are analyzed continuously by the PSS with alarms in the water inventory control station area or control room when measured values go beyond normal limits. -Analog signals are indicated on recorders and/or l

processed in.the computer.

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3The pect-20 ident re pling =yeten vill be ured te perf:em-at 1:::t n:nthly ::::ter :: 1:nt :- ;1: :::1 ::: f:: gr- : i::

7 tcpec, chierid:, ::nductivity, pH, : ygen, and hydrogen.V For l

tr:inin; 2nd :p:::Sility t : ting, ev:ry 5 :: nth: : dilut:d i

liquid gr:b ::nple vill b: d::en, tr:::perted, and :::ly::d fee-heren in the het 125. iThie :- ple vill be '-=dled er a pert-seeident highly rediesetive cenple. '# In rdditi:n, gvery 6 i

l months a; containment air sample will be analyzed for hydrogen, 9,$5 oxygen, and gamma isotopic.

Classroom training will also be provided on system operation and the proper handling of highly l

radioactive samples.

f l

The Suppression Pool, RHR-A and RHR-B shall be sampled through the_ Post Accident Sample System separately _in consecutive six-4 I

month intervals, rotating sampling personnel fot training 5

purposes, such that all three points are sampled on an l

18-month interval.

l 7.7.1.12 Startuo Transient Monitorina System i

{

A startup transient monitoring system will provide recordings for selected parameters during startup and warranty testing.

l This system will remain as a permanent plant monitoring system j

upon completion of the startup test program.

e All permanent cables and equipment will be installed to the same standards as all other plant equipment as described in Section 8.3.1.2.3.

All temporary cabling will be associated only with instrumenta-l tion which is also temporary.

No temporary connections will be made to permanent plant circuitry which is safety related.

All temporary circuitry will be installed in a manner consistent j

with the separation requirements of Regulatory Guide 1.75.

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7.7-62 Rev. 5 12/90 t

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UFSAR i

j 7.5.1.2.17 Meteorological Parameters i

i Meteorological information, as discussed in subsection 2.3.3, I

is available in the control room from the BOP computer.

The i

power supply for meteorological instrumentation is from normal station power.

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7.5.1.2.18 Pgst-Accident Sampling

'The chemical f condition of the reactor coolant may be f

i determined byge-pling via the post-accident sampling system 1

(PASS) nd cheervinghe indic: tion:_On the in;telled "A00 inetr" entatien er by M ecting M b M vi:

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?.'.SS and j

analyzing these samples in the 4raNo 5

"radiotsolgg(g analysis)af *he \\ hot chemistry lab.

reacter ceel?nt, e grsh e==ple rom PAS 9 is analyrEd in the chenistry lib.

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The PASS is powered from BOP station power and can draw samples of reactor coolant end dryr:ll and containment l

atmospheres.

After the results of the chemical and l

radioisotopic analyses are determined, the results can then be communicated to the control room.

(Reference Section 9.3.2)

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j 7.5.1.2.19 Automatic Depressurization System Air Receivers 4

l Two pressure signals are transmitted from two independent i

pressure transmitters to pressure switches which input into j

meters to provide monitoring as shown in Figure 5.2-8.

These

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pressure switches also provide signals to annunciators and BOP

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computers.

The power is supplied from the 125V class 1E dc j

system.

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7.5.1.3 Bvoassed and Inocerable Status Indication i

i Bypasses within the engineered safety feature systems are indicated on the ESF panels by lights and are alarmed by the plant annunciator.

Automatic indication is provided in the control room to inform i

the operator that a system is inoperable.

Annunciation is l

provided to indicate that a system or part.of a system is not j

operable.

For example, the reactor protection (trip) and the containment and reactor vessel isolation control system have

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annunciators lighting and sounding whenever one or more channels of an input variable are bypassed.

Bypassing is not allowed in the trip logic or actuator logic.

Bypasses of certain infrequently used pieces of equipment, such as manual locked open valves, are not automatically annunciated in the control room; however, capability for manual activation of each system level bypass indicator is provided by means of handswitches in the control room for those systems that have these infrequently used bypasses, e'

7.5-16 Rev. 5 12/90 f

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UFSAA TABLE 7.5 2 (Cont.)

POST-ACCIDENT MONITORING INSTRUMENTATION CCNS Oualification Control i

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Measured Type / Cat Environ Seismic GCNS Pot ** r Room SPDS QA i

variable Note 1 Note 4 Note 5 Range SugTy Redundancy Display Note 3 Instrument Vent Flow E/2 N/A.

No Note 20 Not[20 None Note 20 No No Note l5 20 For Fuel Handling Area SCTS & CTNT Vent l

Accident Monitor Turbine Bld E/3 No No Note 18 Non-IE None Note 19 No No Note 18 s.nd Offgas/

Rad Waste Bld Normal &

Accident Range Noble Gas Vent Flow E/3 No No Note 20 Mon-IE None Note 20 No No Note 20 for Turbine R1d and Offges/

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Rad Weste Bld Particu-E/3 No No Note 21 Note 21 None Note 17 No No Note 21 lates Halo-and fans All Note 19 Identified Plant Release Points; Samp-ling w/Onsite Analysis Capa-bility Wind E/3 No No 0-540*

Mon-IE Note 25 Indicatior.

Yee No C84-ZT-N018 Direction

& MO21 Cind E/3 No No 0-90 mph Mon-IE Note 25 Indication Yes No C84-ST-NO19 Speed

& NO22 Estimation E/3 No No

-10*F -

Non-!E Note 25 Indication Yes No CS4-TT-NO23 af Atmos-

+20*F AT

& NO2O pheric 4

ttability Accident Sampling

1. Primary Coolant and Sup-pression Pool Cross E/3 No No 10 pC1/ml to Non-IE None Mone N/A N/A Grab Sample Activity 10 C1/ml*

Camuna E/3 No No Isotopic Mon-1E None Mone N/A N/A Grab Sample Spectrum

Analysis, Boron E/3 No No 0-6000 Mon-IE None None N/A N/A Crab Sample Content ppm *

• Minamum range Sheet 5 of 18 Rev. 5 12/90

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I 00 UFSAR TABLE 7.5-2 (Cont.)

POST *ACC1 DENT MONITORING INSTRUlmrfATION GCNS Qualification Control Meisured Type / Cat Envaron seismic GGNS Power Room SPDS QA V riable Note 1 Note 4 Note S Range Supply Rediiadaaey Display Note 3 Instrument em i 4 u w /s m

m n en sn 7- -

n_,w u,i u,i g,

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sanam u-en u._

o f, u,i rm

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2. Containment Air Hydrogea E/3 No No O to 10%*

Mon =IE None None N/A N/A Crab Samp1 Note 26 Caygen E/3 No No O to 30%*

Non=1R None None N/A N/A Crab Sampi pote 26 Casuna E/3 ido No Isotopic

.Non=15 None None N/A N/A Grab Sampi Spectrum Analysis

• Miniana range o

Sheet 6 of 18 Rev. 2 12/87

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i UFSAR I

9.3.2 Process Sampling System 1

l 9.3.2.1 Design Bases j

9.3.2.1.1 Safety Design Bases 4

l The seismic design and quality group classifications a.

I of sample lines and their components conform to the classification of the system to which they are con-i j

nected up to and including the second isolation valve.

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b.

Sample points located inside the containment terminate at the containment sampling station and do not penetrate i

the containment wall, with the exception of the sample l

points for the post-accident sampling station.

All sampling lines have the process isolation valves i

c.

located as close as practical to the process taps.

i d.

The sampling panels are designed to minimize contami-l nation and radiation at the sample station.

Appro-i priate shielding and area radiation monitors minimize l

radiation effects.

l 9.3.2.1.2 Power Generation Design Bases

]

The process sampling system (PSS) collects representative j

liquid and gas samples for analysis and provides the analytical j

information required to monitor plant and equipment performance 1

and changes to operating parameters.

l The process sampling system is designed to function during all plant operational modes under individual system requirements.

Design guidelines related to PSS capabilities, obtaining representative samples and safety are described in the following-paragraphs and Table 9.3-3.

9.3.2.2

System Description

9.3.2.2.1 General Description The process sampling system provides sampling of all principal fluid process streams as'sociated with plant operation.

The process sampling system consists of the following:

Permanently installed sampling nozzles and sample a.

lines.

Sampling panels with analyzers and associated sampling b.

equipment.

9.3-4 Rev. 3 12/88

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Provisions for local grab sampling.

c.

l 9.3.2.2.2 Sampled Process Streams and Analyzed Parameters i

'a The process streams to be sampled are shown on P& ids with the i

sample point symbols.

The SE symbol is used for Remote Sample i

Points which are. connected with remote sampling panels or i

analyzers.. The SX symbol designates Local Sample Points provided for local crab sampling (see Figures 1.8-1 and 1.8-2).

Figures 9.3-5 through 9.3-8b show the most important 4

sampling panels and condenser leakage detection sampling.

The sample points and local analyzers are shown on the P&ID of individual systems.

For sampling of radioactive gases see section 11.5.

Table 9.3-3 provides a list of sample points, associated P&ID figures, and analyzed parameters.

ProvisionsforObtainingRehresentativeSamples 9.3.2.2.3 The following provisions are incorporated into the PSS:

a.

Where practicable, a sample takeoff connection is located in a turbulent flow zone, where fluid streams are well mixed, after a minimum straight run of 3 pipe diameters of process pipe.

(Where physically possible, a straight run of 10 pipe diameters is preferred.)

b.

The connection is made at the side of horizontal process pipe.

. Sampling nozzles designed for insertion into the c.

streams are provided for process pipes 2-1/2 inches and larger, unless the process of fluid conditions dictate otherwise.

d.

Sampling lines are sized to maintain turbulent flow and to minimize purge time.

Routing is as short and straight as possible.

Tubing with large radius bends is used to avoid traps and dead legs.

e.

Sampling nozzles, lines, and associated valves and fittings are fabricated from stainless steel material.

f.

Heat tracing of sampling lines is provided where necessary to prevent crystallization or solidification of contents.

g.

Sampling equipment is designed for flushing and blowdown in order to remove sediment deposits, air, and gas pockets.

Provisions are made to purge sample lines and to reduce plateout or precipitation in sample lines.

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GG UFSAR i

h.

Provisions are made to sample the bulk volume of j

tanks.

The standby liquid control system storage tank j

will be sampled from the top opening so that any low points and potential sediment traps can be avoided.

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As shown in Table 9.3-3, the post-accident sample system is also capable of drawing and analyzing sampi g of the containment I

and dryr:11 atmosphere.

The.locationd of sample lined, j

as shown in Figure 7.5-5, seg in'relatively open aread of the j

containment with adequate communication.

4 The noncondensible hydrogen and fission product noble gases that will be released to the drywell and containment post-l accident are assumed to form a homogeneous mixture.

The even mixing of noncondensible gases will be promoted by:

Natural convection as a result of temperature gradients a.

i in the drywell and the cascading effect of ECCS water i

exiting from a break l

b.

Turbulence resulting from the operation of containment sprays c.

Depressurization of the reactor coolant system via the l

sequential opening of safety / relief valves distributed around the suppression pool; this will. result in an approximate uniform distribution of noncondensibles in

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the containment.

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d.

Turbulence resulting from the localized burning of j

hydrogen as initiated by the hydrogen igniter system 3

l Based on the above, the dry' dell red containment atmosphere sampleg will be representative of actual conditions.

The system has the capability to draw a grab sample from the containment 2nd the dryw;11 for either on-site or off-site

analysis, i

t A representative-core sample is directly dependent on the amount of mixing of the reactor coolant from the core region

-j with that of the sam'ple location.

Obtaining a sample of reactor coolant that is representative-of core conditions is achieved normally by sampling from the recirculation system via the recirculation loop A sample point as listed in Table 9.3-3.

Therefore, adequate mixing is accomplished by forced circulation provided by the recirculation pumps.

However, core flow circulation in the BWR is inherent without the use of recirculation pumps.

Lack of communication between the downcomer and the core could result in disruption of the major natural circulation flow.

This situation is not likely to occur in jet

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GG UFSAR pump plants because of the open communication between the regions.

In cases where the recirculation pumps are not available, naturally induced coolant flow is established or can be maintained by the density difference between the downcomer region and the core, provided such density difference head is sufficient to balance the losses in the loop.

4 9.3-6a Rev. I 12/86 L

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UFSAR 2

The primary natural circulation loop is between the downcomer j

and the core (see Figure 9.3-30).

Due to boiling in the core i

region, a large difference in densities is available to drive natural circulation flow from the downcomer through the jet

)

pumps and into the shroud region.

The flow due to natural circulation is given in Figure 4.4-5.

l The sample locations for reactor coolant listed ir. Table 9.3-3, assuming that.the recirculation pumps are inoperable, l

would be either RER loop Af RER loop B, Or *'e j et p" r J

-d& M uses.

In the event of an accident, it is estimated that good mixing is achievable in 10 to 20 minutes.

This time is based upon the time required for natural circulated flow to i

complete one internal circulation loop with the flow rate at about 1 percent of rated flow.

The circul: tic: fler Of pririry 00:1:nt by ::tur:1 circul:ti:n is sufficient te provide sirir; ef the prirery ccelant euch that a e-r.le tsken frem th-jet p" r diffuser lecati:n eculd be repreeentative of core conditiens.

Therefore, obtaining a representative sample is dependent upon maintaining natural circulation flow.

First, for an accident such as a DBA LOCA, it is assumed that the majority of the flow of reactor coolant will be the path of least resistance and will be out of the break area and eventually into the suppression pool, with inventory being maintained by ECCS systems.

For this worst case, then, sufficient mixing would still take place in that the entire suppression pool and reactor coolant effec-tively become the same fluid, and a sample from the suppression pool er the jet p"-

would be representative.

Based on the Grand Gulf analysis, it is concluded that provisions exist for adequate mixing of the core and downcomer fluids and that samples taken at the sample points indicated for reactor coolant in Table 9.3-3 will be representative of core conditions.

9.3.2.2.4 Sampling Panels Different process conditions, water quality, and analyzing equipment require special treatment of individual sample streams.sLThese specific requirements are incorporated in the design of the process sampling system whose P& ids are shown in Figures 9.3-5 through 9.3-8b.

5 7 Figure 9.3-5 shows the reactor water sample station and feed-water corrosion product monitor supplied by GE as part of the NSSS.5? urther discussion related to sampling and analysis of F

reactor coolant and the coolant chemistry requirements is provided in subsection 5.2.3.2.2.

9.3-7 Rev. 2 12/87

s GG UFSAR e$ Figures 9.3-6 through 9.3-8a show the individual sampling trains inside the non-NSSS supplied sampling panels. 5A typical high pressure and high temperature sample passes through the inlet panel shut-off valve, through the filter, is rough a

cooled, and pressure reduced to about 25 psig.57The second cooler provides the final temperature conditioning to 77 F before the sample enters the conductivity on pH analyzers.

3 special manual valves are provided to~take grab samples for laboratory analysis.

The sample stream can be monitored by temperature and pressure indicators.f Flow through each analyzer is adjustable by means of the sample flowmeter.I'The train is connected to a coumon header through another flowmeter for continuous blowdown.

-Samples entering the sampling panel at low temperature (below 140 F) require no rough cooling.g Pressure reduction equipment e

6-C O

_O o

n n o am

l r

GG l

UFSAR l

is also provided for initial sample pressure.

For convenience, the sampling panels are divided into several sections, each taking care of special functions:

a.

Pressure and temperature conditioning section.

b.

Grab sampling section with a sink which can be flushed with demineralized water, aample hood, and other equipment for protection against radioactive l

contamination.

c.

Analyzer and monitoring section.

A reclamation header is provided in the turbine building sampling station for all clean water continuous samples to minimize waste processing requirements.

Figure 9.3-7a shows the post-accident sampling station.

Samples are admitted into the system by means of valves which are controlled remotely.

Demineralized water is used to purge the l

system.

Purge control is accomplished remotely.

Coolant samples are cooled by Component Cooling Water with a tube-in-shell type stainless steel sample cooler.

The PASS utilizes grab sampling capability and laboratory instrumentation to monitor the following pathways:

l o

Reactor Coolant (

Dilut:d/ Undiluted) 4 o

Containment Atmosphere (Undiluted)

In-line instrumentation will be used to monitor the following parameters:

Reacter " ster Hydreg:

a 0

Reactor 'dit0r ;M o

Reactor Water Conductivity O

??icter "eter Di 01/ d 0 ys n Containment Atmosphere Oxygen and Hydrogen o

0 Dry':? ell Atre:pher: Oxygen nnd IIydr;; n Grab samples utilize either sample cylinders that have double-shut-off, quick disconnect, and remotely operated valves or a syringe / septum arrangement.

Boron sampling and analysis (frem 2 diluted :nspic) shall be accomplished if the SLC system has been initiated and within 9.3-8 Rev. 9 12/95 yr-,

-,..,,,,_,v

s GG UFSAR i

3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of the time that a decision is made to obtain a i

I post-accident sample, as required by NUREG-0737, Item II.B.3.

The post-accident sampling station is provided with return lines for disposal of the samples; liquid samples are returned to the suppression pool through the RHR system, and the gas samples are returned to the drywell.

The ventilation exhaust from the sampling station area is filtered with a charcoal adsorber and HEPA filters.

The electrically powered components associated with the post-accident sampling system are supplied from a Class 1E MCC.

This MCC is tripped on receipt of an accident signal, but it can be manually reconnected locally.

9.3.2.3 Safety Evaluation a.

The sample nozzles and lines are designed, fabricated, installed, and tested in accordance with the requirements of the process lines from which the samples are taken.

b.

The reactor water sample line (recirculation system) penetrating the drywell wall has two motor-operated isolation valves (one inside and one outside the drywell), which close automatically on an isolation signal from the Containment and Reactor Vessel Isolation Control Systam.

This line is classified as

~

ASME Code,Section III, Class 2.

c.

Reactor coolant to the reactor sample station is sampled with a GE-designed sample nozzle which has an 1/8-inch diameter port hole facing into the flow stream.

This type of sampling nozzle is suitable for obtaining a representative sample and also provides a i

restriction to limit reactor coolant loss from a rupture of the sample line.

d.

Sample nozzles are stress analyzed for the most severe process conditions to avoid failure.

e.

Pressure reduction valves and other devices (pressure regulators and safety relief valves) are provided for protection of operators and/or equipment (refer to Figures 9.3-5 through 9.3-8b).

f.

Reactor water and main steam sample lines are of sufficient length to permit decay of short-lived radionuclides in order to protect sampling personnel.

g.

All sample lines connected to seismic Category I systems are analyzed as seismic Category I lines up to.

and including the second isolation valve; main process 9.3-9 Rev. 9 12/95

{

r oc j

UFSAR j

i pipe code classification is applicable.

Sample lines j

l from the second isolation valve to'the panel / analyzer j

are in conformance with ANSI B31.1, Power Piping Code.

I i

h.

All samples lines have provisions for purging and driining sample streams to an appropriate waste treatment system or to the systems of their origin.

1 9.3.2.4 Tests and Inspection i

The sample nozzles and associated piping, tubing, fittings, and valves are tested and inspected in accordance with the require-ments of the main process pipes from which the samples are taken.

l The process sampling system is proved operable by its use during normal plant operation.

Grab sampling is provided for laboratory analysis for verification of' proper operation and

j..

calibration of continuous analyzers.

)

j 9.3.2.5 Instrumentation Applications The most important water samples are analyzed continuously in the centralized sampling panels where all samples are adjusted to required conditions.

Temperature, pressure, and flow instruments are used to verify and control individual samples.

The analyzers provide alarms when measured values go beyond normal limits.

Analog signals are indicated on recorders and/or processed in the computer.

9.3.3 Floor and Equipment Drainage Systems 9.3.3.1 Design Bases 9.3.3.1.1 Safety Design Bases a.

Drainage from ECCS. equipment rooms is configured to prevent flooding of ECCS equipment via discharge piping backflow.

b.

The design allows for detection of abnormal leakage from emergency core cooling systems and from the dry-well and containment sumps.

c.

Containment and drywell penetrations are fabricated to the requirements of ASME Code,Section III, Class 2.

Auxiliary building penetrations are fabricated to the requirements of ASME Code,Section III, Class 3.

Only these portions of the system have been identified as safety-related and are classified accordingly.

9.3-10 Rev. 0

UFSAR TABLE 9.3-3 (Cont.)

Samole Samole Point Fieure Parameters SG17-N173 RWCU phase sep.

11.2-10 G

tank sampler SG17-N171 Condensate phase 11.2-12a G

j sep, tank (A)

(New Figure) sampler (M-0039X)

'1 SG17-N172 Condensate phase (11.2-12a)

G sep. tank (B)

(New Figure) sampler (M-0039X)

S SG17-N170 Spent resin tank (11.2-12b)

G sampler (M-0039Y)

}

s.

Post-Accident Samole Station at El. 93-0 j

(Turbine Buildinal l

Containment 7.5-5 G, H, 0 2

Atmosphere

-Cry.;;l1 7. 5 -- 5 Gr- %

?.tmo:phero-

-Reoire L^op ?.

5.1-2 0,-H, DO ' C' E" 2

Jct rump Lin O

5. ? 3 0, ",~* DO ' C' E" 2

RHR Loop A 5.4-17 G, H, DO, C,1HF 2

RHR Loop B 5.4-16 G, H, DO, C,14b 2

Suppression Pool 6.2-82 G, H, D0, C, 1ME 2

t.

Solid Radwaste System SG18-N170 Waste holding 11.4-1 G

tank sampler Symbols used under parameters in Table 9.3-3:

C Conductivity Sheet 9 of 10 Rev. 9 12/95

.