ML19261E298: Difference between revisions

From kanterella
Jump to navigation Jump to search
(Created page by program invented by StriderTol)
 
(Created page by program invented by StriderTol)
 
(One intermediate revision by the same user not shown)
Line 16: Line 16:


=Text=
=Text=
{{#Wiki_filter:}}
{{#Wiki_filter:.
              -                      LONG ISLAND LIGHTlNG COM PANY m._
s gffofwap
                        $$                    SHOREHAM NUCLEAR POWER STATION
_  3 4. _ - .. u.      P.O. DOX 610. NORTH COUNTRY RC AD
* WAOING RIVER N.Y.11792 May 16, 1979                                                            SNRC-241 Richard C. Hand, Esq.                                @
Reilly and Like                                              f          %
200 West Babylon Street                          .      59lM Babylon, N. Y. 11702                                            g[
o y      "
                                                                $      ,,,      L        .LM y      <$/ V      Q Mr. Marc W. Goldsmith                                                S Energy Research Group, Inc.                            og g
400-1 Totten Pond Road Waltham, Mass.            02154 Shorehnm Nuclear Pcwer Station - Unit 1 Docket No. 50-322 Gentlemen:
As you k."sw,          representatives from LILCO, Stone & Webster, General Electric, and the NRC Staff met with you on May 2,        1979 to resolve or narrow the contentions addressed in LILCO's first two sets of summary disposition motions, dated December 18, 1978 and February 5, 1979 and additional contentions addressed by certain Staff interrogatories.
During that meeting you asked several questions that either required consultation with personnel who were not present, or you recuested copies of documents that were not available at the meeting. The responses to ycur questions are enclosed.
I understand that the next meeting will be held at the Shoreham construction site Project Cffice on June 6, 1979, and that the purpose of the meeting will be to ccmplete action en the abcVe contentions, prepare a stipulation concerning them, and move on to the following SC contentions:                      Sa(i-xx), 5b (i-111) ,
Sc (iii & iv), lla (1-tii) , and 12a(iii & v-vii) .
Ver          civ yours,
                            )                                              .
                  !-        LW%c                          nc c)} ?.,l \
s      t-P. Sc,'arro,                                                                      6 7907 05 0 d%^ /
Project Manager Shorehan Nuclear Pcwer Station                              cc:    Members of the Board All Parties JPM/cl
 
LILCO RESPONSES TO SC/ ERG QUESTICNS AND REQUESTS
    .              FOR ADDITIONAL INFORMATION FROM MAY 2,  1979 MEETING General Q.1    Provide Shoreham's operational QA nanual.
A.1    A copy of the Shoreham Operating Q:A Manual is herewith forwarded to ERG.
Q.2    Provide a copy of the Shoreham Fire Hazards Analysis Report.
A.1    The bhoreham Fire Hazards Analysis Report was forwarded to ERG on May 3, 1979.
SC Contentions Addressed In First Set of Summary Disposition Motions 7a (vi) 0.1    Provide engineering drawings of the spent fuel pool rack embedments and the associated fuel rack mounting pads.
A.1  Engineering drawings of the spent fuel storage rack floor enbedments and the associated fuel storage rack mounting pads are herewith forwarded to ERG.
Q.2  Provide photographs of the spent fuel rack base plates.
A.2  Photographs of the spent fuel storage rack base plates are h;&ewith forwarded to ERG.
Q.3  What is the maximum rate of leakage to the spent fuel pool leakage collection system before scme corrective action must be taken? What action will be taken to locate the source of leakage and correct it?
A.3  The leakage collection system is a means to detect and collect any leakage through the liner seam welds and the welds in the spent fuel storage rack embedments. Because these welds are thoroughly tested during fabricaticn and installation, they are not anticipated to leak. The probability of these welds leaking is so small that most other plants do not have leakage collection systems for their spent fuel pcols.              ,, c, ; ^7
 
There are a total of 13 leak test zones: four (4) each for the spent fuel pocl, the reactor head cavity, and the dryer and separator storage pool, as well as one for the spent fuel storage rack embedments in the spent fuel pool, Each test zone is piped to a separate test station.
If leakage were to occur, the corrcctive action would depend on the severity of the leakage. The action would be either an intensified monitoring program or a program to locate and repair the leak. To locate the leak, probably compressed air would be introduced into the leakage collection system for the appropriate zone. The compressed air will pass through the leak and appear as bubbles in the spent fuel pool witer, thus indicating the location of the leak.
Then, depending on the location of the leak, a variety of techniques could be used to repair the leak.
p  f l
                                                                ''l w
(. I t
 
-                        SC Contentions Addressed in Second Set of Swnmary Disposition Motions 2 0 a ( i) - (ii) 0.1    Provide a copy of the S&W radiological waste treatment systems maintTinability study.          Explain the general context of the study and provide representative examples of the ger.eral mcdifications made to improve piant maintainability.
A.1    A copy of the radwaste building portion of the Shoreham Maintainability Study is herewith forwarded to ERG.
This study discussed, in general terms, how the maintainability of equipment could be improved and the potential radiation dose levels decreased.          As a result of this study, many pipes and valves were relocated and other design modifications were implemented to improve plant maintainability and to reduce occupational exposure.        For example:
: 1. Relocation of pipi      in the anion, cation, and regeneration rs        orage tank area to allow for accessa i      , to pumps and the ultrasonic resin cleant..
: 2.  'iajor revisions of piping and valves in the radwaste filter demineralizer area to allow for accessibility to valves with a minimum of exposure to plant personnel.
: 3. Addition of platforms and relocation of air operated valves in the waste collector, ficor drain, and discharge waste sample tank area.
: 4. Relecation of non-radicactive equipment and addition of shield walls to reduce maintenance exposure for the solidificatio. system.
: 5. Open areas were established for valkways, as well as for equipment access, removal, and laydown.
2 h)  2
 
Q.2  Provide a summary and/or minutes of meetings between
-          LILCO and S&W regarding industrial experience and Shoreham design considerations for ALARA and maintainability of radiological waste solidification systems.
A.2  The following are provided herewith to ERG in response to this request:
: 1. Notes of Conference Nos. 690 and 708
: 2. SNPS-2036E letter dated 3/11/76
: 3. S&W letter to ATCOR dated 11/16/76
: 4. LIL-lll24 letter dated 8/04/77
: 5. LIL-13993 letter dated 2/26/79
: 6. SNPS-2881E letter dated 3/22/79 Q.3  Provide any Commonwealth Edison documentation on the Dresden Unit 1 decontamination that could be useful to Shoreham. Do these documents suggest that changes should be made at Shoreham, and if so, are they being made?
A.3  There are very substantial differences in the design concepts of the Dresden Unit 1 Nuclear Generating Station as compared to Shoreham. Therefore, it is doubtful that the Dresden Unit 1 decontamination experience will be applicable to Shoreham. At this time, LILCO is attempting to cbtain documentation regarding the Dresden decontamination.
If, after reviewing this documentation, LILCO has determined that it is relevant to Shoreham, LILCO will supply a copy to ERG.
2 /k
                                          ~    ~
                                                                / c j
 
SC Contentions Based on Certain Staff Interrogatories 6a(i) 0.1    Will LILC0 monitor the          esults of LaSalle's future feedwater nozzle inspectivn?
A.1    Yes, L'T      ' will ronitor.  :hrough G.E. and NRC reports, the resu;.ts cc f e r-  ections of feedwater nozzles at plants, such as La3.lle, e.:P nozzle designs similar to Shorehams.
Q.2    Will G. . fve available for inspection at the next informal summary - c;osition meeting a copy of G.E. report NEDE-21480, BWR Fee'.wster Fo le/Sparger Interim Program Report, dated February, 1977?
A.2    Yec, a copy of the above requested report will be available for ERG's perusal at the next informal summary disposition meeting, currently scheduled for June 6, 1979.
8a (i) - (ii) 0.1    Provide a copy of EPRI report NP-309SY, Human Factors Review of Nuclear Power Plant Control Room Design, dated November, 1976.
A.1    The requested report is presertly out of print.          Ecwever, arr1ngements were made with the EPRI report center on May 3, 1979 to forward a copy of the summary directly to ERG. In a telephone conversation between M. Goldsmith (ERG) and J. Morin (LILCO) on May 4, 1979, Mr. Goldsmith agreed to use S&W's copy of the full report en a loan basis.      S&W's copy was forwarded to ERG on May 4, 1979.
Q.2    Provide G.E. internal design specifications for centrol rocm anels.
A.2    The G.E. internal design specifications for control rocm panels are herewith forwarded to ERG.
2)g    /$ b
 
Q.3 What criteria were used by LILCO in customizing the Shoreham control panels?
A.3 During the years 1970 and 1971, LILCO worked with both General Electric and Stone & Webster to develop the Main Control Room NSSS and Balance of Plant Control Board layouts. In order to enhance the arrangement of the control panels and to establish a philosophy consistent with those presently in use at other LILCO power stations, members of the LILCO Electric Production Department visited other nuclear power stations including Oyster _ Creek, Vermont Yankee, and Dresden to look at their main control room panels and to learn how those panels could be mproved. As a result, LILCO established certain criteria for the design of Shoreham's control panels. These criteria included the following:
: 1. The general arrangement of the control panels would be functionally similar to the General Electric simulator located in Morris, Illinois to facilitate the transfer of operator training experience to Shoreham's control pane ~.s.
: 2. Mimics for the systems would be presented on the control panels.
: 3. The mimics would be color-coded to distinguish between the various systems.
: 4. The size of the mimic presentations would be selected to distinguish main process lines frcm auxiliary flow paths.
: 5. Process controllers would be included in the system mimic and indicators showing the parameters affected wculd be in close proximity to the associated controllers to create cohesive functional groups.
: 6. The maximum height of the upper row cf gauges wculd be approximately 5'4" above ficar level uad the maximum height of the recorders would be 4'4" above floor level. This would enhance the ability to read and maintain the meters and recorders.
: 7. The control panel annunciators would be functionally grouped with their corresponding systems.
29 7 Og
      .            _      _. ____-6  _._    ._
: 8. Certain annunciators wculd be color-ccded to distinguish the more important functions; for example, RPS scram or ECCS Inop alarms would be color-coded red and ECCS system degradation alarms would be color-coded yellow.
: 9. The systems would be separated by a space in the panel or a vertical line to distinguish individual systems.
: 10. Control switches for motor operated valves would be color-coded to indicate whether they were throttleable or non-throttleable type valves.
: 11. The range switches for the intermediate range monitors would be color-coded to correspond to the recorder pens they were associated with to aid the operator in range switching during reactor heatups and shutdowns.
: 12. Both sound powered phones and plant interecm system phones would be provided along the length of the control panels to enhance operator communications with support personnel in the plant.
: 13. The operator's desk would be positioned in front of and facing the reactor board portion of the main control panel.
: 14. Flat-faced gauges were selected to reduce a glare problem which had been experienced as a result of the use of curved-faced gauges.
: 15. The Control Rocm will be carpeted to keep the general area noise level to a minimum.
: 16. Ccmputer cutput peripherals were integrated into the control board to give additional flexibility to the operator in the information available to him. The peripheral devices incorporaced included CRT's, ccmputer trend recorders, and ccmputel digital displays.
: 17. Annunciater panel alarms were given different tone signals to assist the cperator in identifying the area of the control board from which the alarm was originating.
: 18. A consistent chilosophy on the use of indicating lights was established, and bnplcmented within the control board presentation, i.e., red -
running, open; green - stop, closed, etc.
2f)1 2  3
 
Q.4    Provide a representative sample of control panel mimics (drawings and photographs).
A.4    A representative specification drawing and control rocm photographs that illustrate typical control panel mimics, controls, and monitors are herewith forwarded to ERG.
Q.5    Provide a description of operator training and requalification program.
A.5    The Shoreham Training Program and Retraining Program are discussed in detail in Shoreham FSAR 55 13.2.1 and 13.2.2, respectively.
13 a (ii) , (v) 01    What were Shoreham's earlier water chemistry limits?      What are Shoreham's current water chemistry limits?
A.1    The early and current water quality requirements are provided in the following table:
Early Maximum    Current Maximum Reactor Water:                  Parameter              Limits              Limits Normal operation              Conductivity          1.0 umhos/cm Reactor pressurized                                                  11.0 umhos/cm Chlorides            0.2 ppm max.      10.2 ppm max.
pH                    5.6 - 8.6          5.6 - 8.6 Short duration Reactor pressurized          Max. conductivity      10 umhos/cm        10 umhos/cm Max. time above        2 weeks /yr.        2 weeks /yr.
1 umho/yr.
Max. chloride          1.0 ppm            0.5 ppm Max. time above        2 wk/yr.            2 wk/yr.
0.2 ppm Reactor Depressurized        Conductivity          10 umhos/cm        10 umhos/cm Chloride              0.5 ppm            0.5 ppm pH                    6.0 - 8.5          5.3 - 8.6
                                                                ') n 1    ;'- 1 L if    I-
 
13a (ii) , (v) (Con't.)
Early Maximum      current Maximum Feedwater:                  Parameter          Limits                Limits During normal full        Conductivity    4 0.1 umho/cm          10.1 umho/cm Power operation            Chlorides      ~1. ppb                42 ppb pH                6.5 - 7.5              6.5 - 7.5 Metallic        <30 ppb              615 ppb Impurity        <2 ppb copper        62 ppb During initial            .dtallic'
                                            <lb0ppbaverage        <50 ppb startup lst 500 hours              Impurity        4500 ppb peak          ~100 ppb peak
_9-
                                                        ')a (1-]- (, 0
                        ..}}

Latest revision as of 21:41, 1 February 2020

Forwards Responses to Questions Raised During 790502 Meeting of Nrc,Util,S&W & GE Re Narrowing of Util Contentions Addressed in NRC Interrogatories.Next Meeting Will Be Held 790606
ML19261E298
Person / Time
Site: Shoreham File:Long Island Lighting Company icon.png
Issue date: 05/16/1979
From: Novarro J
LONG ISLAND LIGHTING CO.
To: Goldsmith M, Hand R
EXIENE, REILLY & LIKE
References
SNRC-241, NUDOCS 7907050524
Download: ML19261E298 (10)
v • d • e


Text

.

- LONG ISLAND LIGHTlNG COM PANY m._

s gffofwap

$$ SHOREHAM NUCLEAR POWER STATION

_ 3 4. _ - .. u. P.O. DOX 610. NORTH COUNTRY RC AD

  • WAOING RIVER N.Y.11792 May 16, 1979 SNRC-241 Richard C. Hand, Esq. @

Reilly and Like f  %

200 West Babylon Street . 59lM Babylon, N. Y. 11702 g[

o y "

$ ,,, L .LM y <$/ V Q Mr. Marc W. Goldsmith S Energy Research Group, Inc. og g

400-1 Totten Pond Road Waltham, Mass. 02154 Shorehnm Nuclear Pcwer Station - Unit 1 Docket No. 50-322 Gentlemen:

As you k."sw, representatives from LILCO, Stone & Webster, General Electric, and the NRC Staff met with you on May 2, 1979 to resolve or narrow the contentions addressed in LILCO's first two sets of summary disposition motions, dated December 18, 1978 and February 5, 1979 and additional contentions addressed by certain Staff interrogatories.

During that meeting you asked several questions that either required consultation with personnel who were not present, or you recuested copies of documents that were not available at the meeting. The responses to ycur questions are enclosed.

I understand that the next meeting will be held at the Shoreham construction site Project Cffice on June 6, 1979, and that the purpose of the meeting will be to ccmplete action en the abcVe contentions, prepare a stipulation concerning them, and move on to the following SC contentions: Sa(i-xx), 5b (i-111) ,

Sc (iii & iv), lla (1-tii) , and 12a(iii & v-vii) .

Ver civ yours,

) .

!- LW%c nc c)} ?.,l \

s t-P. Sc,'arro, 6 7907 05 0 d%^ /

Project Manager Shorehan Nuclear Pcwer Station cc: Members of the Board All Parties JPM/cl

LILCO RESPONSES TO SC/ ERG QUESTICNS AND REQUESTS

. FOR ADDITIONAL INFORMATION FROM MAY 2, 1979 MEETING General Q.1 Provide Shoreham's operational QA nanual.

A.1 A copy of the Shoreham Operating Q:A Manual is herewith forwarded to ERG.

Q.2 Provide a copy of the Shoreham Fire Hazards Analysis Report.

A.1 The bhoreham Fire Hazards Analysis Report was forwarded to ERG on May 3, 1979.

SC Contentions Addressed In First Set of Summary Disposition Motions 7a (vi) 0.1 Provide engineering drawings of the spent fuel pool rack embedments and the associated fuel rack mounting pads.

A.1 Engineering drawings of the spent fuel storage rack floor enbedments and the associated fuel storage rack mounting pads are herewith forwarded to ERG.

Q.2 Provide photographs of the spent fuel rack base plates.

A.2 Photographs of the spent fuel storage rack base plates are h;&ewith forwarded to ERG.

Q.3 What is the maximum rate of leakage to the spent fuel pool leakage collection system before scme corrective action must be taken? What action will be taken to locate the source of leakage and correct it?

A.3 The leakage collection system is a means to detect and collect any leakage through the liner seam welds and the welds in the spent fuel storage rack embedments. Because these welds are thoroughly tested during fabricaticn and installation, they are not anticipated to leak. The probability of these welds leaking is so small that most other plants do not have leakage collection systems for their spent fuel pcols. ,, c, ; ^7

There are a total of 13 leak test zones: four (4) each for the spent fuel pocl, the reactor head cavity, and the dryer and separator storage pool, as well as one for the spent fuel storage rack embedments in the spent fuel pool, Each test zone is piped to a separate test station.

If leakage were to occur, the corrcctive action would depend on the severity of the leakage. The action would be either an intensified monitoring program or a program to locate and repair the leak. To locate the leak, probably compressed air would be introduced into the leakage collection system for the appropriate zone. The compressed air will pass through the leak and appear as bubbles in the spent fuel pool witer, thus indicating the location of the leak.

Then, depending on the location of the leak, a variety of techniques could be used to repair the leak.

p f l

l w

(. I t

- SC Contentions Addressed in Second Set of Swnmary Disposition Motions 2 0 a ( i) - (ii) 0.1 Provide a copy of the S&W radiological waste treatment systems maintTinability study. Explain the general context of the study and provide representative examples of the ger.eral mcdifications made to improve piant maintainability.

A.1 A copy of the radwaste building portion of the Shoreham Maintainability Study is herewith forwarded to ERG.

This study discussed, in general terms, how the maintainability of equipment could be improved and the potential radiation dose levels decreased. As a result of this study, many pipes and valves were relocated and other design modifications were implemented to improve plant maintainability and to reduce occupational exposure. For example:

1. Relocation of pipi in the anion, cation, and regeneration rs orage tank area to allow for accessa i , to pumps and the ultrasonic resin cleant..
2. 'iajor revisions of piping and valves in the radwaste filter demineralizer area to allow for accessibility to valves with a minimum of exposure to plant personnel.
3. Addition of platforms and relocation of air operated valves in the waste collector, ficor drain, and discharge waste sample tank area.
4. Relecation of non-radicactive equipment and addition of shield walls to reduce maintenance exposure for the solidificatio. system.
5. Open areas were established for valkways, as well as for equipment access, removal, and laydown.

2 h) 2

Q.2 Provide a summary and/or minutes of meetings between

- LILCO and S&W regarding industrial experience and Shoreham design considerations for ALARA and maintainability of radiological waste solidification systems.

A.2 The following are provided herewith to ERG in response to this request:

1. Notes of Conference Nos. 690 and 708
2. SNPS-2036E letter dated 3/11/76
3. S&W letter to ATCOR dated 11/16/76
4. LIL-lll24 letter dated 8/04/77
5. LIL-13993 letter dated 2/26/79
6. SNPS-2881E letter dated 3/22/79 Q.3 Provide any Commonwealth Edison documentation on the Dresden Unit 1 decontamination that could be useful to Shoreham. Do these documents suggest that changes should be made at Shoreham, and if so, are they being made?

A.3 There are very substantial differences in the design concepts of the Dresden Unit 1 Nuclear Generating Station as compared to Shoreham. Therefore, it is doubtful that the Dresden Unit 1 decontamination experience will be applicable to Shoreham. At this time, LILCO is attempting to cbtain documentation regarding the Dresden decontamination.

If, after reviewing this documentation, LILCO has determined that it is relevant to Shoreham, LILCO will supply a copy to ERG.

2 /k

~ ~

/ c j

SC Contentions Based on Certain Staff Interrogatories 6a(i) 0.1 Will LILC0 monitor the esults of LaSalle's future feedwater nozzle inspectivn?

A.1 Yes, L'T ' will ronitor. :hrough G.E. and NRC reports, the resu;.ts cc f e r- ections of feedwater nozzles at plants, such as La3.lle, e.:P nozzle designs similar to Shorehams.

Q.2 Will G. . fve available for inspection at the next informal summary - c;osition meeting a copy of G.E. report NEDE-21480, BWR Fee'.wster Fo le/Sparger Interim Program Report, dated February, 1977?

A.2 Yec, a copy of the above requested report will be available for ERG's perusal at the next informal summary disposition meeting, currently scheduled for June 6, 1979.

8a (i) - (ii) 0.1 Provide a copy of EPRI report NP-309SY, Human Factors Review of Nuclear Power Plant Control Room Design, dated November, 1976.

A.1 The requested report is presertly out of print. Ecwever, arr1ngements were made with the EPRI report center on May 3, 1979 to forward a copy of the summary directly to ERG. In a telephone conversation between M. Goldsmith (ERG) and J. Morin (LILCO) on May 4, 1979, Mr. Goldsmith agreed to use S&W's copy of the full report en a loan basis. S&W's copy was forwarded to ERG on May 4, 1979.

Q.2 Provide G.E. internal design specifications for centrol rocm anels.

A.2 The G.E. internal design specifications for control rocm panels are herewith forwarded to ERG.

2)g /$ b

Q.3 What criteria were used by LILCO in customizing the Shoreham control panels?

A.3 During the years 1970 and 1971, LILCO worked with both General Electric and Stone & Webster to develop the Main Control Room NSSS and Balance of Plant Control Board layouts. In order to enhance the arrangement of the control panels and to establish a philosophy consistent with those presently in use at other LILCO power stations, members of the LILCO Electric Production Department visited other nuclear power stations including Oyster _ Creek, Vermont Yankee, and Dresden to look at their main control room panels and to learn how those panels could be mproved. As a result, LILCO established certain criteria for the design of Shoreham's control panels. These criteria included the following:

1. The general arrangement of the control panels would be functionally similar to the General Electric simulator located in Morris, Illinois to facilitate the transfer of operator training experience to Shoreham's control pane ~.s.
2. Mimics for the systems would be presented on the control panels.
3. The mimics would be color-coded to distinguish between the various systems.
4. The size of the mimic presentations would be selected to distinguish main process lines frcm auxiliary flow paths.
5. Process controllers would be included in the system mimic and indicators showing the parameters affected wculd be in close proximity to the associated controllers to create cohesive functional groups.
6. The maximum height of the upper row cf gauges wculd be approximately 5'4" above ficar level uad the maximum height of the recorders would be 4'4" above floor level. This would enhance the ability to read and maintain the meters and recorders.
7. The control panel annunciators would be functionally grouped with their corresponding systems.

29 7 Og

. _ _. ____-6 _._ ._

8. Certain annunciators wculd be color-ccded to distinguish the more important functions; for example, RPS scram or ECCS Inop alarms would be color-coded red and ECCS system degradation alarms would be color-coded yellow.
9. The systems would be separated by a space in the panel or a vertical line to distinguish individual systems.
10. Control switches for motor operated valves would be color-coded to indicate whether they were throttleable or non-throttleable type valves.
11. The range switches for the intermediate range monitors would be color-coded to correspond to the recorder pens they were associated with to aid the operator in range switching during reactor heatups and shutdowns.
12. Both sound powered phones and plant interecm system phones would be provided along the length of the control panels to enhance operator communications with support personnel in the plant.
13. The operator's desk would be positioned in front of and facing the reactor board portion of the main control panel.
14. Flat-faced gauges were selected to reduce a glare problem which had been experienced as a result of the use of curved-faced gauges.
15. The Control Rocm will be carpeted to keep the general area noise level to a minimum.
16. Ccmputer cutput peripherals were integrated into the control board to give additional flexibility to the operator in the information available to him. The peripheral devices incorporaced included CRT's, ccmputer trend recorders, and ccmputel digital displays.
17. Annunciater panel alarms were given different tone signals to assist the cperator in identifying the area of the control board from which the alarm was originating.
18. A consistent chilosophy on the use of indicating lights was established, and bnplcmented within the control board presentation, i.e., red -

running, open; green - stop, closed, etc.

2f)1 2 3

Q.4 Provide a representative sample of control panel mimics (drawings and photographs).

A.4 A representative specification drawing and control rocm photographs that illustrate typical control panel mimics, controls, and monitors are herewith forwarded to ERG.

Q.5 Provide a description of operator training and requalification program.

A.5 The Shoreham Training Program and Retraining Program are discussed in detail in Shoreham FSAR 55 13.2.1 and 13.2.2, respectively.

13 a (ii) , (v) 01 What were Shoreham's earlier water chemistry limits? What are Shoreham's current water chemistry limits?

A.1 The early and current water quality requirements are provided in the following table:

Early Maximum Current Maximum Reactor Water: Parameter Limits Limits Normal operation Conductivity 1.0 umhos/cm Reactor pressurized 11.0 umhos/cm Chlorides 0.2 ppm max. 10.2 ppm max.

pH 5.6 - 8.6 5.6 - 8.6 Short duration Reactor pressurized Max. conductivity 10 umhos/cm 10 umhos/cm Max. time above 2 weeks /yr. 2 weeks /yr.

1 umho/yr.

Max. chloride 1.0 ppm 0.5 ppm Max. time above 2 wk/yr. 2 wk/yr.

0.2 ppm Reactor Depressurized Conductivity 10 umhos/cm 10 umhos/cm Chloride 0.5 ppm 0.5 ppm pH 6.0 - 8.5 5.3 - 8.6

') n 1  ;'- 1 L if I-

13a (ii) , (v) (Con't.)

Early Maximum current Maximum Feedwater: Parameter Limits Limits During normal full Conductivity 4 0.1 umho/cm 10.1 umho/cm Power operation Chlorides ~1. ppb 42 ppb pH 6.5 - 7.5 6.5 - 7.5 Metallic <30 ppb 615 ppb Impurity <2 ppb copper 62 ppb During initial .dtallic'

<lb0ppbaverage <50 ppb startup lst 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> Impurity 4500 ppb peak ~100 ppb peak

_9-

')a (1-]- (, 0

..

Retrieved from "https://kanterella.com/w/index.php?title=ML19261E298&oldid=2382192"