ML20140B304

Rev 2 to Removal of Ice Condenser Vent Rod & Containment Spray/E5 Water Heat Exchanger Monitoring Sys

ML20140B304
Person / Time
Site:	Oyster Creek
Issue date:	08/16/1995
From:	Buechel S GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20140B292	List: ML20140B287 ML20140B304
References
000661-017, 000661-017-R02, 661-17, 661-17-R2, NUDOCS 9706060134
Download: ML20140B304 (9)

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' Oyster Creek Nuclear Generating Station Safety Evaluation 000661 - 017 l

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13 J o u 3 ENasdaar. Tcettnical Functions D6HCW?@ OMWIL-Safety / Environmental Determination and 50.59 Review l.

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Documem No. (/f spoucebie'l U Doc.Rev.No. V '

SE No. Oml.cj.7 Type of Activity (modince p, ure, test, expedrei, or afocurnent):

l 1. Does this document invoMe any potential non-nuclear environmemal concern? O Yes @ No f To answer this quescon, review the Environmental Determination (ED) form. Any YES answer on the i

ED form requires an Environmental Impact Assessment by Environmental Controls, po'r j

1000-ADM-4500.03. If in doubt, consult Environmental Comrois or hi;rc,s rstal Ucensing for assistance, If all answers are NO, further environmental review is not required. In any event, continue with Question 2 below.

i l 23 is this activity / document listed Section i or 11 of the matnces in Corporate Procedure @ Yes

. 1000-ADM-1291.017 O No

' If the answer to quescon 1 is NO, stop here. This procedure is not maphe=Ma and no docume.mtion is required. (if this actmty/documem is Ested in Secnon IV of 1000-ADM-1291 review on a case-by-case

basis to determine applicabirrty.) if the answer is YES, proceed to question 3.

3. Is this a new actmty/ document or a substantive revision to an activity / document?

j Eyes O No (See Exhibit 2, paragraph 3, this procedure for examples of non-substantive changes.)

j If the answer to question 3 is NO, stop here and complete the approval secnon below. This procedure is not applicable and no documentation is required. If the answer is YES, proceed to answer all remaining j questions. These answers become the Safety / Environmental Determinanon and 50.59 Review.

4 Does this activity / document have the potential to adversely affect nuclear safety

, or safe plant operations? S Yes [ N'o l 5. Does this activity / document require revision of the system /companent he,h

! Eyes O No in theofFSAR part the SAR7 or otherwise require revision of the Technical Sef,c.;T.c ;;cas or any other i

} 6. Does the activity / document require revision of any procedural or operating desenanon.

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'. lyes .SNo in the FSAR or otherwise require revision of the Technical Fpecifications or any other part of the SAR7 j

5

7. Are tests or experimems conducted which are not described in the FSAR, the Technical Specifications or any part of the SAR7 O Yes  % No -

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IF ANY OF THE ANSWERS TO QUESTIONS 4. 5,6, OR 7 ARE YES, PREPARE A WRITTEN SAFETY EVALUATION FORM.

i l If the answers to 4,5,6, and 7 are NO, this produdes the occurrence of an Unroviewed Safety Quescon or i

Techrdcal Specifications chance. Provide a written statement in the space provided below (use back of shat if necessary) to suppan the determinsoon, and list the documems you checked.

NO. because: _

Documents checked: OcffAA II.M.3 mows 6541.isc e 5.15d-st-ou_ -

8. An the design criteria as outEned in TMI 1 SDD T1-000 Div. I or DC-SDD-000 Div. I Plant Level Critena affected by, or do they affect the activity / document?

O Yes BNo If YES, indicate how resolved:

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! Answs' the fotowing quesdons and padde reason (s$ ter each answer per Exhibit 7. A simple

! mesment of concluelon in heen is 'not sunisient. The anope and depe of each reason should be l i commensurme with the esisty signulmenos and compisulty of the pmposed ohenge.

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1. wm implemerannon of the aanvitpiocument advensey snoct numiser l

! .. sesmy or sais pient operanons? , DWs SNo l l

yhe sonowing questions comprise the same consideradens and i evalumaan a doenrmine N an Unmviewed Saluty Quesden suisis:

2. is the pmbebaty of ocounence or the annesquences of an anoident or manuneen of equirnent imponent so ashly paviously maiussed in the Sahty Analysis Report increased? CWs $No a is its posseisty for an accident or messunamon of a dmerent type than .

any evaluated previously in the Salsty Analysis Report cremed? Oyes IDNo

4. is the margin of aslety as defined in the basis for any Dohnical Speamcenon mduced? DWs $No .

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If any answer above is "yes" en impant on nucieer asisty or an Unesviewed Safety Queellon*

eden N an adverse impact on nucieer asisty ades revlee or redesign. If an unreviewed asis.

ty quashon with no adverso impact on numiser seisty exists tensord to Lloonsing with any ad.

duional documentation to support a request for NRC appmel prior to implementing appeal,

s. speoNy whether or not any of the tonoung are rogured, and W "yes" indicate how it was resolved x 4 < Yan TF No 4 .

s. Does the actMtpdocument require X an update of the F3AR?

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SAFETY EVALUATION CONTINUATION SHEET Page 4 of 8 SE-000661-017 Rev. No. 2 l i

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TABLE OF CONTENTS I I

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1.0 PURPOSE...................................................................................................................................5 )

2. 0 S YS TE M S A FFE CTE D ............................ . .............................................................. .................... 5 3.0 E FFE CT S O N S A FE TY ....................................... .................... ............ ........... ... ...... .......... 6  ;

4.0 CONCLUSION

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4 l SAFETY EVALUATION CONTINUATION SHEET  ;

Page S of 8 i SE-000661-017 i Rev. No. 2 4

j. 1.0 PURPOSE ,

1 l He purpose of this safety evaluation is to evaluate the removal of the fbilowing two systems, Isolation l Condenser Vent Radiation Monitoring System and Containment Spray / Emergency Service Water Heat l1-Exchanger Radiation Monitoring System, from the plant configuration as these monitoring systems presents the potential for operator confusion.

The scope of the modification for the Isolation Condenser Vent Radiation Monitoring System involves

, disconnection and physical removal of log ratemeters RN0007A3, 4, 5 and 6 from Control Room ,

. Panel 2R and replacement with blank panels, disconnection of four (4) input signals to Recorder l l RN006B on Control Room Panel 10F, disconnection of signal circuitry to Alarm Window 10F-1-e (A l %.

. Vent Hi) and to Alarm Window 10F-2-e (B vent Hi) on Panel 10F and replacement with twd (2) blank

wmdows, and physical removal of all associated abandoned wiring within tle Control Room.

j Detectors RE-RN0004A3,4,5 and 6 and their associated shielding and cabling are to be physically .

removed from Reactor Building Elevation 95'3" l 1

s 1

, The scope of the modification for the Contamment Spray / Emergency Service Water Heat Exchanger l

Monitoring System involves disconnection and physical removal of log ratemeters RN0040Al, 2, 3 i and 4 from Control Room Panel 2R and replacement with blanks, disconnection of four (4) input signals to Recorder RN006B on Control Panel 10F, disconnection ofinput signals to Alarm Window 10F-4-g (Area / Vent /Effl. Dnscl.) on Panel 10F, disconnection of signal circuitry to Alarm Window 10F-1-g (ESW A/B Hi) and to Alarm Window 10F-2-g (ESW C/D Hi) on Panel 10F and replacement with two (2) blank windows, and physical removal of all associated abandoned wirmg within the Control Room. Detectors RN0038Al,2,3 rd 4 and their associated shielding are to be abandoned in place at Reactor Building Elevation 23'6" 2.0 SYSTEMS AFFECTED 2.1 - ne plant systems affected by this modification consist of the following:

l 2.1.1 Radiation Monitoring System (661)

' l 2.1.2 Main Control Room Panels (61!) ,

2.1.3 Plant Annunciator System (616) ,

2.1.4 Emergency Condenser System (Isolation Condenser)(211) 2, 2.1.5 Containment Spray System / Emergency Service Water System (241/532) 2.2 The systems affected by this modification are shown on the following baseline drawings:

2.2.1 Process Radiation Monitor System, GE Dwg No. 846D686, Sh.1, Rev.18.

2.3 The affected systems are described in Section 6.2,6.3 and 11.5.2.3 of the Oyster Creek FS AR. 'L-008/264

SAFETY EVALUATION CONTINUATION SHEET Page 6 of 8 SE-000661-017 i

Rev. No. 2 3.0 EFFECTS ON SAFETY 3.1 Will implementation of the activity / document adversely affect nuclear safety or safe plant ,

operations?

4 No. The existence of the Isolation Condenser Vent Radiation Monitoring System and the Contamment Spray / Emergency Service Water Heat Exchanger Radiation Monitoring System ' 2.

=

represents a potential source of operator confusion since background radiation in the sicinity of the monitors masks their capability for leak detection. Existing surveillance procedures muumize the potential for loss of integrity during normal operating and emergency conditions.

Although original plant design took credit for vent radiation monitors for leak detection

capability, alternate means are available to detect leakage in a timely manner.

. The isolation condenser vent radiation monitors were intended to provide a leak detection j function and not an effluent quantification function (Ref. memo 5350-91-180, dated 8/21/91).

l Tube wall integrity is verified on an ongoing basis by surveillance requirements, including 4 tracking of temperature and level every four hours. With respect to leak detection capability, the Technical Specifications require isolation condenser integrity, but do not specify the use of radiation monitors for verification. _

in order to increase the availability of the Isolation Condensers (ICs) as heat sinks, NUREG 0737 Item II.K.3.14 required that their isolation on high radiation be switched from the Main

Steam Line Radiation Monitors to the IC Vent Radiation Monitors. In response, GPUN, by a 2-l letter from Ivan R. Finfrock to the NRC, dated April 30,1981 (copy attached), clarified the fact that Oyster Creek did not isolate the ICs on high radiation. Isolation is provided by detection of excessive flow in the steam line to and condensate line from the IC In its response letter, dated December 18, 1981, the NRC stated that the GPUN response to item II.K.3.14 was acceptable and the item was considered resolved.

The Containment Spray / Emergency Service Water Heat Exchanger Radiation Monitoring [2.

System was part of the original plant operation. Similarly, the detectors do not adequately

. perform their design function because background radiation in the vicinity of the monitors masks their effectiveness for leak detection. The Containment Spray /ESW heat exchanger does not function during normal plant operation, when the tube side is fed with service water.

The system is fed with Emergency Service Water dming monthly surveillance testing and following an accident.

The tubes are not subject to significant temperature or pressure excursions during post-LOCA conditions. The tube side is normally maintained at a positive differential pressure with respect to the shell side (exception identified in LER 84-26). Should the pressure differential drop below 5 psid (indication of a possible leak) the Control Room is alarmed. On a monthly basis, surveillance testing is performed which includes verification of tube and shell ride pressures and differential pressure indication.

Leakage from the torus into the ESW through the heat exchangers during non accident conditions has negligible off site dose consequences.

008/264

SAFETY EVALUATION CONTINUATION SHEET Page 7 of 8 SE-000661-017 Rev. No. 2 Per Rad engineering the contanunation concentration of the torus water will not be permitted to increase significantly from current average concentrations during operation because higher levels of contanunation present operational problems in terms of contamination of plant areas and personnel dose. Given this upperbound to the contammation of the torus water Rad Engineering calculated, Memo 6632-96-001, (See references) that a leak rate of 10,000 gpm through the ESW heat exchangers would be necessary to exceed 10CFR50 Appendix I limits.

The maximum annual dose to an individual in an unrestricted area per 10CFR50 Appendix I is specified as 5 mR.10CFR20.1301 has a less restrictive annual dose limit to an individual in an unrestricted area of 100 mR. During normal operation the toms water may be at a one psi greater pressure than the ESW water. Memo 5310-96-011 (See references) concluded that a very conservative torus to ESW leak rate is I gpm. Given the Rad Engineering calculation that a 1 gpm leak rate has the potential to produce an off site dose of 1/10000 of Appendix I , 2.

limits a 1 gpm leak rate has negligible off site consequences.

Catastrophic failure of a heat exchanger tube is not a credible event because, per drawing PX-D9361-S-100, the heat exchanger is tubed with titanium, rated for operation at 250 psi and subjected to only a 1 pr.i differential pressure (Ref. Memo 5310-96-036, dated 2/16/96).

Nevertheless an order of magnitude esumate was made of the consequences of a catastrophic tube failure. The 18 foot 3/4 inch tube is assumed to experience a guillotine break 3 feet from one end. The ends of the tube deflect so that both ends discharge freeF Flow is estimated using Crane Technical Paper 410 Appendix B page 14, Flow of Water Through Schedule 40 Steel Pipe. The use of a pipe table to estimate flow from tubing is conservative because 3/4 inch Sch 40 pipe has a larger inner diameter than 3/4 inch tubing. Given a 1 psi differential pressure the total leak rate can be estimated to be less than 30 gpm. A 30 gpm leak rate has the potential to produces an off site dose of 3/1000 of Appendix I limits. This has negligible off site consequences. From these cases it may be concluded that the leakage of torus water into the ESW through the heat exchanger tubes including the case of catastrophic tube failure has negligible off site dose consequences.

The system is subject to leak testing during refueling ' outages, with an acceptance criterion of -

zero leakage. The heat exchanger tubes are of titanium which is resistant to corrosion. 1 Shell side (torus water) chemistry analysis is performed on a monthly basis (Ref. Memo 5350-91-022, dated 4/9/91). -

3.2 Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report increased?

No. Removal of these radiation monitors has no impact on the operation of any plant system l

and as such has no impact on the probability or consequences of any previously evaluated , j event.  !

The radiation monitors are not relied upon to perform a post-accident function. j i

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008/264 l l

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} SAFETY EVALUATION CONTINUATION SHEET I

, . Page 8 of 8 SE-000661-017 Rev. No. 2

! 3.3 is the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report created?

No. These radiation monitors do not affect operation of any plant systems, and as such their removal will not create the possibility for an accident or malfunction of a different type.

3.4 Is the margin of safety as defined in the basis for any Tech Spec reduced? -

No. The contamment spray / emergency service water radiation monitors are not discussed in the Technical Specifications. The isolation condenser vent radiation monitors are discussed in l the basis for Technical Specifications Section 3.8. as ". . provided to alert the operator of a tube leak in the isolation condenser . . ". As described in Section 3.1 above, the background radiation in the vicinity of these monitors, while the isolation condensert are inservice, masks their capability for leak detection. However, Oyster Creek procedures would lead to manual isolation in the event of a tube leak, usir.g a combination of on-site monitoring for radioactivity 2.

and control room indications and alarms, such as steam line temperatures and shell levels and temperature.

Based on the above, it is concluded that the removal of these radiation monitors has no impact on any margin of safety as defined in Technical Specifications.

The basis for Technical Specification Section 3.8 will be revised to delete the description of isolation condenser vent radiation monitors.

3.5 This activity will require revision of Design Basis Document for '661 System and the following FSAR Sections and Tables:

+ Sections 11.5.2.3,6.2.2.2,6.3.1.1.2,6.3.1.1.3,1.9.31

+ Tables 11.5-1, 6.2-10 2.

4.0 CONCLUSION

From the justifications provided above, it is concluded that the proposed changes will not have any I

adverse effect on plant safety, and do not represent an unreviewed Safety Question. The changes will however require a change to the Oyster Creek FSAR as described in Section 3.5.

2.

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I mNuclear GPU Nuclear

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201 263-6500 7ELEX 136-482 Writer's Direct Osa! Number; a

April 30, 19El .

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I Otractor, .

Of fice of Nuclear Reactor Regulation ,] 4,\l; ' jc U. S. Nuclear Regulatory Commission ,

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Washington, D. C. 20555 Deer Sir:

Subject:

Oyster Creek Nuclear Generating Station Docket No. 50-219 l

NUREG 0737, Item iI.K.3.14 1

I NUREG 0737, item II.K.3.14 specifies a design change to the isolation 1

J

' condenser system Isolation logic. The description as provided in the NtREG is incorrect for the Oyster Creek station. The isolation condenser system utilized at the Oyster Creek Nuclear f acility does not isolate upon a high-radiation i signal in the mein steam line lead ng to the isolation condensers. The Oyster t

system utllizes excessive flow In the steam lines to and condensate lines from l i  ;

the isolation condensers as the only isolation signal for the system isolation valves.

i Because the Oyster Creek system is different from the one described in the <

l' NUREG, 9e feel no additional modifications are required to meet the objective of this NUREG item.

I f you should have any questions concerning this matter, please call Mr.

, James Knubel (201-455a8753).

Very truly yours, Ivan R. Fin Jr. '

Vice Prest t ir .s

, . ;s GPU Nuclear is a part of the General Public Utilities System TOTAL P.Cr?

QS[GJSS