ML13331A793
| ML13331A793 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 04/09/1985 |
| From: | SOUTHERN CALIFORNIA EDISON CO. |
| To: | |
| Shared Package | |
| ML13331A791 | List: |
| References | |
| NUDOCS 8504110326 | |
| Download: ML13331A793 (25) | |
Text
DESCRIPTION OF PROPOSED CHANGE NO. 150 AND SAFETY ANALYSIS PROVISIONAL OPERATING LICENSE DPR-13 This is a request to revise Provisional Operating License DPR-13, License Condition 3.E, to allow operation of San Onofre Unit 1 until the November 1985 refueling outage without the requirement to perform an interim steam generator inspection.
Description License Condition 3.E, Steam Generator Inspections, was revised by the issuance of NRC amendment No. 80 to Provisional Operating License No. DPR-13, to require an interim steam generator inspection within 6 Equivalent Full Power Months (EFPM) of operation from the backfit outage that began on February 27, 1982. This interim inspection was required since previous information submitted to the NRC staff had not provided sufficient justification for allowing longer periods of operation between steam generator inspections. In order to present additional information justifying a longer steam generator inspection interval, meetings were held between the NRC staff and SCE on February 27 and March 14, 1985, and a report entitled "1985 Reevaluation of Steam Generator Inspection Interval, March 1985 (March 1985 report)," was provided to the NRC by SCE letter dated March 19, 1985.
The purpose of the report Is to document re-evaluation of the basis for the current steam generator inspection interval and provide sufficient information to justify a longer steam generator interval.
The enclosed proposed change to License Condition 3.E revises the current requirement for an interim steam generator inspection prior to the November 1985 refueling outage. The revision of this condition is based upon the re-evaluated Intergranular Attack (IGA) degradation rate that was derived using the methodology in the March 1985 report. The re-evaluated IGA rate (10% growth per* 15 EFPM) justifies the resumption of a refueling cycle interval (15 EFPM) for steam generator inspections.
The refueling cycle steam generator inspection interval is required by Appendix A Technical Specification 4.16, Inservice Inspection of Steam Generator Tubing. The next such inspection is scheduled for the refueling outage which will begin on November 30, 1985, after approximately 10.5 EFPM of operation, assuming optimum operation, since the last inspection. It is expected that the steam generator inspection in the Fall 1985 refueling outage will serve to confirm the conservatism of re-evaluated IGA degradation rate established in the March 1985 report. The confirmation of the conservatism of the re-evaluated IGA degradation rate (10% per 15 EFPM) will also confirm the appropriateness of deleting the requirement for future interim steam generator inspections.
Existing License Condition See Enclosure 1 Proposed License Condition See Enclosure 2 504110326 80409 P(
05000206 PDR?
-2 Safety Evaluation The Proposed Change as discussed above shall be deemed to constitute a significant hazards consideration if positive findings are made in any of the following areas:
I. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No This proposed change is a revision to an administrative constraint on the operation of the facility. The previous limitation on operating interval was based upon a limited review of eddy current data from 39 steam generator tubes, known to be in areas of the highest IGA activity and in the deepest part of the steam generator sludge pile, which has been determined to be a contributor to high IGA activity. The re-evaluation presented in the March 1985 report demonstrated that the steam generator tubes that were within the 1980 sleeving repair boundary showed eddy current signatures that are distinctly different than those eddy current signatures in the nonsleeved tubes.
The review of eddy current data from steam generator inspections prior to 1980 also demonstrated that while eddy current signatures of tubes that are within the sleeving boundary and in areas of highest IGA activity were undergoing distinct changes due to increasing IGA degradation, the eddy current signatures of the nonsleeved tubes showed evidence that the IGA, that may be present, has been relatively dormant over the period of review. The conclusion is that the "actual" IGA degradation rate in the nonsleeved tubes is negligible.
The revised steam generator water chemistry controls implemented after the 1980 Sleeving Repair Project, later revised to incorporate the guidance of the EPRI PWR Secondary Water Chemistry Guidelines, provide assurance of the continued maintenance of conditions not conducive to the propagation of IGA.
Notwithstanding the perceived no-growth rate in the nonsleeved tubes, the March 1985 report derived a conservative degradation rate for the purposes of establishing an appropriate operating interval between steam generator inspections.
The 10% growth rate per 15 EFPM, if a current condition of 40% IGA is considered, results in the.conclusion that at the end of 15 EFPM of operation, 50% of the tube wall would be remaining. Using a 10% allowance for eddy current examination error, the expected remaining minimum tube wall would be 40%. The results of the analysis discussed in the conclusions of the March 1985 report conclude that a 40% tube wall is sufficient to withstand the worst case postulated accident loads.
Therefore, it is concluded that since the assumed IGA growth rate regarding the current nonsleeved tube condition and the allowances for error are conservative, the steam generator inspection interval of 15 EFPM is appropriate and conservative, and operation of San Onofre Unit 1 until the Fall 1985 refueling outage, in accordance with this Proposed Change, will not involve a significant increase in the probability or consequences of an accident previously evaluated.
-3 II. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No The degradation of steam generator tubes could lead to a steam generator tube rupture which is a previously analyzed accident for San Onofre Unit 1. Additionally, the operating interval between steam generator inspections is a previously analyzed condition and the reevaluated IGA degradation of 10% growth per 15 EFPM is consistent with this analysis. Therefore, operation of the facility in accordance with this Proposed Change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
III. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?
Response: No The re-evaluated IGA degradation rate in the March 1985 report is based upon the review of eddy current inspection data from previous inspections that demonstrate its appropriateness for the nonsleeved steam generator tubes.
Assurance of the continued validity of the re-evaluated IGA degradation rate is provided by the revised steam generator water chemistry controls now in place and the SCE management commitment to the control of steam generator water chemistry. Therefore, operation of San Onofre Unit 1 in accordance with this Proposed Change does not involve a reduction in a margin of safety.
This Proposed Change will allow operation of San Onofre Unit 1 until the Fall 1985 refueling outage, when a steam generator inspection will be required.
The extension of the current operation restriction is justified based upon the completion of a review of eddy current examination data to assess the current condition of the non-sleeved steam generator tubes at San Onofre Unit 1 and to assign a limiting IGA degradation rate to those tubes, for the purposes of determining an appropriate steam generator inspection interval.
Therefore, this proposed change is similar to example (iv) of amendments not likely to involve a significant hazards consideration published in 48 FR 14864 dated April 6, 1983.
Safety and Significant Hazards Determination Based on the Safety Analysis, it is concluded that (1) the Proposed Change does not involve a significant hazards consideration as defined by 10 CFR 50.92; and (2) there is reasonable assurance that the health and safety of the public will not be endangered by the Proposed Change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Environmental Statement. - Existing License Condition -
Proposed License Condition
ENCLOSURE 1 EXISTING LICENSE CONDITION
L. This license applies to the San Onofre Nuclear Generating Station Unit No. 1 (hereinafter, the "facility"), a pressurized light water reactor and associated steam generators and electric generating equipment. The facility is located on the site of Edison and San Diego near the northern boundary of Camp Pendleton in San Diego County, California, and is described in license application Amendment No. 8 "Final Engineering Report and Safety Analysis," Volumes I, II, and III, and in Supplements 1, 2 and 3 thereto (Amendment Nos. 10, 11, and 13 to the license application).
Said "Final Engineering Report and Safety Analysis" in Amendment No. 8, as supplemented and amended, is hereinafter referred to as the hazards summary report.
This page included FOR REFERENCE ONLY
- 2. Subject to the conditions and requirements incorporated herein, the Atomic Energy Commission (the Commission) hereby licenses:
A. Pursuant to Section 104b of the Atomic Energy Act of 1954, as amended, (hereinafter, the "Act"), Edison, San Diego, Bechtel and Westinghouse to acquire and possess title to the facility as a utilization facil ity, and Edison and San Diego to acquire and possess sole title to the facility at such time as the legal title of Bechtel and Westinghouse shall pass to Edison and San Diego.
B. Edison and San Diego, with Edison acting for itself and as agent for San Diego:
(1) To possess, use and operate the facility as a utilization facility, pursuant to Section 104b of the Atomic Energy Act of 1954, as amended.
(2) Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report (docketed August 7, 1970), as supplemented and amended; (3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation, radiation monitoring equipment calibration and reactor coolant system boron concentra tion measurement, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFRParts 30 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the opertion of the facility.
This page included FOR REFERENCE ONLY
- 3. This license shall be deemed to contain and is subject to the conditions as specified in the following Commission regulations (Title 10, CFR, Chapter 1):
Part 20, Sections 50.54 and 50.59 of Part 50, Section 70.32 of Part 70, Section 40.41 of Part 40 and Section 30.34 of Part 30; is subject to all applicable provisions of the Act and rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:
A. Maximum Power Level Edison is authorized to operate the reactor at steady state power levels up to a maximum of 1347 megawatts thermal.
B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 88, are hereby incorporated in the license. Southern California Edison Company shall operate the facility in accordance with the Technical Specifications.
C.
Fuel Shipments shall not be resumed (except during cold shutdown condition of the reactor) until the modifications described in the licensee's March 21, 1976 report entitled "Spent Fuel Shipping Cask Handling, San Onofre Nuclear Generating Station, Unit 1," and the initial turbine deck load bearing test and visual inspection have been satisfactorily completed.
D. The facility may be modified by implementing the "Sphere Enclosure Project":
as described in Amendment 52 to the Final Safety Analysis for the San Onofre Nuclear Generating Station, Unit 1, submitted December 3, 1975; Supplement to the Sphere Enclosure Project Report, submitted March 1, 1976; Second Supplement to the Sphere Enclosure Report submitted March 25, 1978; additional information submitted by letter dated March 25, 1976 (withheld from public disclosure pursuant to 10 CFR Part 2, Section 2.790(d)).
E. Steam Generator Inspections Southern California Edison shall bring the reactor to a cold shutdown condition to perform an inspection of the steam generators within six equivalent months of -operation from the start of operation from the backfitting outage that commenced on February 27, 1982. The inspection program shall be submitted to the Commission at least 45 days prior to the scheduled shutdown.
Commission approval shall be obtained before resuming power operation following this inspection.
F.
Deleted
ENCLOSURE 2 PROPOSED LICENSE CONDITION
- 3. This license shall be deemed to contain and is subject to the conditions as specified in the following Commission regulations (Title 10, CFR, Chapter 1):
Part 20, Sections 50.54 and 50.59 of Part 50, Section 70.32 of Part 70, Section 40.41 of Part 40 and Section 30.34 of Part 30; is subject to all applicable provisions of the Act and rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:
A.
Maximum Power Level Edison is authorized to operate the reactor at steady state power levels up to a maximum of 1347 megawatts thermal.
B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 88, are hereby incorporated in the license. Southern California Edison Company shall operate the facility in accordance with the Technical Specifications.
C.
Fuel Shipments shall not be resumed (except during cold shutdown condition of the reactor) until the modifications described in the licensee's March 21, 1976 report entitled "Spent Fuel Shipping Cask Handling, San Onofre Nuclear Generating Station, Unit 1," and the initial turbine deck load bearing test and visual inspection have been satisfactorily completed.
D. The facility may be modified by implementing the "Sphere Enclosure Project":
as described in Amendment 52 to the Final Safety Analysis for the San Onofre Nuclear Generating Station, Unit 1, submitted December 3, 1975; Supplement to the Sphere Enclosure Project Report, submitted March 1, 1976; Second Supplement to the Sphere Enclosure Report submitted March 25, 1978; additional information submitted by letter dated March 25, 1976 (withheld from public disclosure pursuant to 10 CFR Part 2, Section 2.790(d)).
E.
Steam Generator Inspections During the refueling outage scheduled to begin no later than November 30, 1985, Southern California Edison shall perform an inspection of the steam generators. The ins.pection program shall be submitted to the Commission at least 45 days prior to the scheduled shutdown. Commission approval shall be obtained before resuming power operation following this inspection.
F. Deleted
NUCCEAR GENERATION SITE CHEMISTRY PROCE RE SO123-III-2.1.1 UNITS 1, 2 AND 3 REVISION 2 PAGE 1 OF 16 EFFECTIVE DATE Off 3 1 UNIT 1 STEAM GENERATOR AND CONDENSATE/FEEDWATER CHEMISTRY CONTROL AND SAMPLING FREQUENCIES TABLE OF CONTENTS SECTION PAGE 1.0 OBJECTIVES 2
2.0 REFERENCES
RECEIVED CDM 2
3.0 PREREQUISITE CC
.31 ;924 2
4.0 PRECAUTIONS SITE FILE COPY 2
5.0 CHECKLIST(S) 3 6.0 PROCEDURE 3
7.0 RECORDS 15 ATTACHMENTS 1 In-line Monitors of Steam Generator, Condensate and Feedwater 16 QA PROGRAM AFFECTING 0189c
NUCLEAR GENERATION SITE CHEMISTRY PROCEDURE S0123-III-2.1.1 UNITS 1, 2 AND 3 REVISION 2 PAGE 2 OF 16 UNIT 1 STEAM GENERATOR AND CONDENSATE/FEEDWATER CHEMISTRY CONTROL AND SAMPLING FREQUENCIES 1.0 OBJECTIVES 1.1 To establish a program of parameter limits and routine sampling requirements for Steam Generator, Condensate and Feedwater chemistry control at Unit 1 in order to inhibit Steam Generator tube degradations.
1.2 To satisfy the requirements of Reference 2.1.1 and proceduralize the guidelines of References 2.3.1 and 2.3.2.
2.0 REFERENCES
2.1 Licensing Commitments 2.1.1 San Onofre Unit 1 Technical Specifications 2.2 Procedures 2.2.1 S0123-III-2.8.1, Unit 1 Secondary System Chemistry Control During Hot Soak, Heat-Up, and Initial Power Escalation Following an Extended Outage 2.2.2 S0123-III-0.4.4, "Communication of Chemistry Conditions via Chemistry Memos" 2.2.3 S0123-III-2.22.1, "Unit 1 Steam-Generator Leak Rate N
Determination" 2.3 Other 2.3.1 Steam Side Water Chemistry Control Specifications, Westinghouse publication 2.3.2 PWR Secondary Water Chemistry Guidelines, EPR1-NP-2704-SR 3.0 PREREQUISITE 3.1 Verify this procedure is the most current revision against a white control file and review all applicable Temporary Change Notices (TCNs).
4.0 PRECAUTIONS 4.1 This procedure is written to be followed in a step-by-step manner.
If a step cannot be performed in accordance with this procedure or, if unexpected results are obtained, contact Chemistry Supervision immediately.
NUCLEAR GENERATION SITE CHEMISTRY PROCEDURE S0123-III-2.1.1 UNITS 1, 2 AND 3 REVISION 2 PAGE 3 OF 16 4.0 PRECAUTIONS (Continued) 4.1.1 Since more than one analysis is addressed in this procedure, each analysis may be performed separately and analyses may be performed in any order as long as the outcome of the procedure is not invalidated.
4.2 This procedure is for use at Unit 1 only.
4.3 Samples may be radioactive and shall be handled in a manner to minimize exposure and to prevent spreading contamination.
4.4 Check the applicable radiation and contamination survey information and use this information to maintain your exposure ALARA. Observe the appropriate Radiation Protection precautions as specified on the Radiation Exposure Permit.
4.5 Observe safety precautions when handling hazardous chemicals.
5.0 CHECKLIST(S) 5.1 None 6.0 PROCEDURE (Unit 1 Only) 6.1 Chemistry Specifications and Sampling Frequencies 6.1.1 Since the program's intent is to adequately maintain water quality, it is necessary for the Supervisor of Unit Chemistry to have the flexibility to change the Chemistry Program in response to any ongoing chemical condition. Therefore, the frequency, limits and parameters analyzed may be altered by the Supervisor of Unit Chemistry under the following guidelines:
.1 Issuance of a memo that cancels automatically after 3 days;
.2 Issuance of a memo with Supervisor of Nuclear Plant Chemistry approval that cancels automatically after 7 days;
.2.1 Any condition that exists after 7 days will require a TCN to the appropriate procedure.
.3 The flexibility provided by this paragraph DOES NOT authorize any relaxation of the frequency, limits and parameters required by the Technical Specifications.
NUCLEAR GENERATION SITE CHEMISTRY PROCEDURE S0123-III-2.1.1 UNITS 1, 2 AND 3 REVISION 2 PAGE 4 OF 16 6.0 PROCEDURE (Unit 1 Only) (Continued) 6.1.2 The specifications stated by this procedure are divided into two groups:
Normal Range and Abnormal Range. The first group of chemistry conditions provides high quality chemistry control while retaining operational flexibility. The abnormal range permits limited full power operation while the cause of an out-of-normal parameter is identified and corrected. If the abnormal range is exceeded, power must be reduced to decrease available superheat in the steam generator tube sheet and, thus, prevent accelerated steam generator corrosion while the affected component is isolated and/or repaired. Immediate shutdown values have been established to prevent continued operation in an agressive corrosion condition which can result in serious damage. Steam generator operation in compliance with these specifications will provide a more reliable overall secondary system with fewer repairs due to long-term corrosion problems.
6.1.3 Operation within the normal range is consistent with long-term system reliability. When the normal range is exceeded, immediate investigation of the problem must be initiated using the appropriate guidelines, sampling frequency increased or conduct a generator blowdown.
The problem must be corrected and the parameter(s) returned to the normal range within one week. If no abnormal range is listed, Chemistry/Station management should be consulted to provide course of action guidance.
6.1.4 When the abnormal range is exceeded, power should be reduced to less than 30% to minimize the feed rate.
Continued plant operation is then possible while corrective action is taken.
Power reduction should be initiated within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of exceeding the abnormal range. The problem must be corrected and the parameter(s) returned to the normal range within one hundred (100) hours. If this cannot be done, Station management should be consulted to provide course of action guidance.
6.1.5 When an immediate shutdown is recommended, the unit should be shut down to at least hot standby and the Steam Generator's makeup water supplied by the auxiliary feedwater system to prevent rapid steam generator corrosion.
6.1.6 The time periods to correct an out-of-normal or out-of abnormal range condition discussed in paragraphs 6.1.3 through 6.1.5 above do not apply to pH control or parameters affected by saltwater leaks.
NUCLEAR GENERATION SITE CHEMISTRY PROCEDURE S0123-III-2.1.1 UNITS 1, 2 AND 3 REVISION 2 PAGE 5 OF 16 6.0 PROCEDURE (Unit 1 Only) (Continued) 6.1.6.1 Chemical additions or Operations actions required to maintain pH within prescribed limits must be taken immediately upon determining a high or low pH condition exists. It is unacceptable to operate several hours with pH out of limits, and this procedure is not intended to imply that remaining outside the pH range for a week is satisfactory.
.2 A saltwater leak can result in a special case of out-of-range condition for Chlorides:(Cl ). The severity of the event may not allow the maintaining of steady state plant operation without some immediate action to minimize the effects of undesirable chemical species. A saltwater leak is confirmed by comparing the Cl and Na in the hotwells. After the leak has been located and isolated, the severity of the effects of the leak need to be evaluated. Refer to Section 6.6 of this procedure for Chemistry action.
6.1.7 The sampling frequencies are designed to provide sufficient analyses to ensure that the secondary systems are operating within the established guidelines of the procedure and to provide adequate checks on continuous monitors. All in-line monitors/recorders when in service shall be recorded whenever the corresponding bench check is performed and as follows:
Modes 1 and 2 -
Every 4 hr Modes 3 and 4 -
Every 8 hrs Modes 5 and 6 -
Every 12 hrs 6.1.8 In-line monitors should be observed frequently by the responsible Nuclear Chemistry Technician. is a list of all in-line monitors. As a minimum, the NCT should check these monitors every 30 minutes when in Modes 1 and 2. He should verify that the equipment is operating properly (both the measuring instrument and the recorder). He should understand or attempt to understand the reasons for any changes observed in recorded parameters. (Operations Department personnel are responsible for checking the Steam Generator blowdown radiation monitor recorder in the Control Room at a frequency specified by Operations.)
6.1.9 When proceeding from Mode 5 to Mode 1, the chemistry limits of this procedure for the next Mode should be attained prior to proceeding to the next desired Mode. If a Mode change is made for some reason other than progression to power operation, Chemistry Supervision should evaluate water chemistry limits and provide recommendations.
NUCLEAR GENERATION SITE CHEMISTRY PROCEDURE S0123-III-2.1.1 UNITS 1, 2 AND 3 REVISION 2 PAGE 6 OF 16 6.0 PROCEDURE (Unit 1 Only) (Continued) 6.1.10 For Modes 1, 2, 3 and 4, suspended iron analysis should be conducted on Condensate and Feedwater for 2-3 days after a mode change or until the analysis result is less than 20 ppb. At that time, the filtration test should be terminated and the wet iron analysis should be started on condensate and feedwater systems once per week.
6.2 Acceptance Criteria 6.2.1 Chemistry parameters found to be out of the normal range on first analysis should immediately be followed up with a second sampling if they are not consistent with other values or in-line monitors and reported in accordance with Reference 2.2.2, except for saltwater leaks, which shall be reported immediately.
6.2.2 Chemistry parameters verified out of the specified range shall be reported to the applicable Chemistry Foreman and the Unit Chemistry Supervisor. Any unusual or significant out-of specification condition should be verbally forwarded to the Shift Superintendent immediately. Recommendations for corrective actions shall be forwarded with the out-of-specification memo.
6.2.3 If an in-line monitor reading is out of the normal range or if an alarm occurs, follow the appropriate guidelines for corrective action.
6.2.4 On occasion, a sample will be taken that appears abnormal in terms of color, clarity, undissolved material (either: floating, suspended or precipitated) and/or gas content (i.e., dissolved gases in liquid samples). When such a sample is taken, in addition to informing supervisory personnel, label and retain the abnormal sample for comprehensive evaluation.
6.2.5 Weekly Blowdown
.1 A maximum (2 1/2 turns) blowdown shall be performed at least once a week, during power operation (Mode 1), to reduce the amount of sludge in the steam generators.
(Normally - approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)
NUCLEAR GENERATION SITE CHEMISTRY PROCEDURE S0123-III-2.1.1 UNITS 1, 2 AND 3 REVISION 2 PAGE 7 OF 16 6.0 PROCEDURE (Unit 1 Only) (Continued) 6.3 PLANT OPERATION (MODES 1 and 2) 6.3.1 Condensate and Feedwater, (e)
NORMAL ABNORMAL PARAMETER FREQUENCY RANGE RANGE pH @ 25*C 0
8.8-9.4
<8.8; >9.4 Chloride D
<50 ppb
>50 ppb
- Hydrazine D
3x0, not to exceed 50 ppb
- Cat Cond 0
<0.5 uS/cm
>0.5 Dissolved Oxygen D
<10 ppb
>10 ppb Conductivity D
<6.5 PS/cm
- Ammonia D
Z1.0 ppm
>1.0 ppm
- Iron W
<20 ppb
>20 ppb
- Copper W
<10 ppb
>10 ppb
- Tritium 3/W
<1x1O-'vCi/ml
>1x10-'vCi/ml, (a) l 6.3.2 Steam Generator Blowdown NORMAL ABNORMAL PARAMETER FREQUENCY RANGE RANGE pH @ 25cC 0
9.4-10.2
<9.4, >10.2,
(
Conductivity D
<125 uS/cm
>125 PS/cm Phosphate 0
15-30 ppm
<15, >30 Chloride 0
<500 ppb
>500 ppb, (b)
Silica D
<300 ppb
>300 ppb Na/PO D
2.3-2.6
<2.3, >2.6, (c)
Gross Activity M,W,F 1-131 Dose Equivalent M
<0.1 uCi/gm
>0.1 uCi/gm (d)
(a) Notify Chemistry Supervision, refer to Reference 2.2.3.
(b) Recommend a unit shutdown if chloride concentration is
>2,000 ppb.
(c) Determine ratio using Marcy-Halstead and analytical methods for comparison.
(d) If DE Iodine is >0.1 vCi/gm, the unit shall be in cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (T.S. 3.4.2).
(e) For Condensate and Feedwater sampling, the designations are as follows:
no asterisk = Run on both Condensate and Feedwater
- = Run on Condensate only
- = Run on Feedwater only
NUCLEAR GENERATION SITE CHEMISTRY PROCEDURE SO123-III-2.1.1 UNITS 1, 2 AND 3 REVISION 2 PAGE 8 OF 16 6.0 PROCEDURE (Unit I Only) (Continued) 6.3 PLANT OPERATION (MODES 1 and 2)
(Continued) 6.3.3 Main Steam NORMAL PARAMETER FREQUENCY RANGE Sodium 3/W
<45 ppb, (a)
Chloride W
Ta)
Carryover M
<0.25%
Cation Conductivity W
70.3, (a)
(a) Compare results of bench analysis with in-line instrument reading.
6.3.4 Hot Well Quadrants NORMAL PARAMETER FREQUENCY RANGE Chloride (a)
<50 ppb (a)
As required and when specified by Chemistry Supervision
NUCLEAR GENERATION SITE CHEMISTRY PROCEDURE S0123-III-2.1.1 UNITS 1, 2 AND 3 REVISION 2 PAGE 9 OF 16 6.0 PROCEDURE (Unit 1 Only) (Continued) 6.4 HOT STANDBY (MODES 3 AND 4), (c) 6.4.1 Condensate and Feedwater NORMAL CORRECTIVE PARAMETER FREQUENCY RANGE ACTION pH @ 250C M,W,F 8.8-9.4 (a)
Conductivity M,W,F
<6.5 pS/cm (a) 4 Cation Conductivity M,W,F
<10pS/cm (f)
Chloride M,W,F
<100 ppb (a)
Hydrazine M,W,F 10-50 ppb or 3x01 (b)
Suspended Iron M,W,F
<500 ppb (a)
Ammonia M,WF
<1.0 ppm (a)
Copper W
<50 ppb (a) 6.4.2 Steam Generator NORMAL CORRECTIVE PARAMETER FREQUENCY RANGE ACTION pH @ 250C D
9.4-10.5 (a)
Conductivity D
<250 uS/cm (a)
Phosphate D
10-60 ppm (a)
Chloride 0
<500 ppb (a)
Na/PO.
0 2.3-2.6 (a), (d)
Tritium M,W,F
<1x10-4 ci/ml Gross Activity M,W,F 1-131 Dose Equivalent M
<0.1 VCi/gm (e)
<500 ppb (a) When parameters are out of the normal range, chemical additions or system dilution by draining and refilling (or feeding and bleeding) should be initiated to restore chemistry within the normal range.
(b) Whichever is greater.
(c) When progressing toward Mode 1 after a refueling or extended outage, review Reference 2.2.1.
(d) Determine ratio using Marcy-Halstead and analytical methods for comparison. Take corrective action when ratio is <1.0 or
>3.0.
(e) If DE Iodine is >0.1 uCi/gm, the unit shall be in cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (T.S. 3.4.2).
(f) Perform on condensate only
NUCLEAR GENERATION SITE CHEMISTRY PROCEDURE S0123-III-2.1.1 UNITS 1, 2 AND 3 REVISION 2 PAGE 10 OF 16 6.0 PROCEDURE (Continued) 6.5 COLD SHUTDOWN (MODES 5 AND 6) 6.5.1 Condensate and Feedwater, (a)
NORMAL CORRECTIVE PARAMETER FREQUENCY RANGE ACTION pH @ 256C 3/W 8.8-9.9 (b)
Conductivity 3/W
<50 uS/cm (b)
Hydrazine 3/W 25-50 ppm (b)
Chloride W
<1000 ppb (b)
Ammonia W
Z1000 ppb (f
Copper M
Z100 ppb (f)
Iron M
Z1000 ppb (f) 6.5.2 Steam Generators (Tave <200*F), (a),
(e)
NORMAL CORRECTIVE PARAMETER FREQUENCY RANGE ACTION pH @ 25cC 3/W 9.3-10.5 (b)
Conductivity 3/W
<160 uS/cm Drain/fill (c)
Chloride 3/W Z1000 ppb Drain/fill (c)
Phosphate 3/W Z50 ppm Drain/fill (c)
Na/PO.
3/W 1.0-3.0 Drain/fill (c)
Hydrazine 3/W 50-150 ppm Drain/fill (c)
Nitrogen Overpressure 3/W
>0 psig (d)
(a) Samples may not be available during periods of dry layup, system maintenance and stagnant conditions.
See Section 6.5.3.
(b) If a parameter is out of the Normal Range, drain and refill, feed and bleed or recirculate or sparge with Nitrogen as directed by Chemistry Supervision. When adding hydrazine to the Cond/FW system, normally no more than one liter should be added to each train.
(c) Steam generators should be drained and filled under nitrogen pressure.
(d) The nitrogen overpressure should be maintained at all times when the steam generators are not on auxiliary feedwater flow.
(e) After make-up water and/or hydrazine is added to a steam generator, recirculation or nitrogen sparging should be performed. A sample should be taken and analyzed for pH, conductivity, chloride, phosphate, sodium and hydrazine within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of recirculation or sparging.
(f) Data is for trending analysis only; corrective action only when directed by Chemistry Supervision.
NUCLEAR.GENERATION SITE CHEMISTRY PROCEDURE SO123-III-2.1. 1 UNITS 1, 2 AND 3 REVISION 2 PAGE 11 OF 16 6.0 PROCEDURE (Unit 1 Only) (Continued) 6.5.3 Steam Generators (Dry Layup)
.1 During periods where it is necessary to place the secondary side of the steam generators in dry layup to allow access for maintenance or other reasons, the chemistry recommendations of 6.5.2 will be held in abeyance until the completion of 6.5.3.2. To prepare the steam generators for personnel access, the following recommendations should be provided to Operations:
.1.1 Immediate Access Drain the generators cold (<1300F).
.1.2 Delayed Access If the generators will not be entered for several days, drain the affected generator cold (<130'F) under nitrogen pressure.
.1.2.1 Maintain the nitrogen until just prior to the desired access, then purge the generator with warm and dry filtered air.
CAUTION The steam generators should be considered
==
=
a confined space and, therefore, the appropriate safety precautions should be taken to ensure a breathable atmosphere exists inside the generator prior to allowing personnel access.
.2 When access to the steam generator is no longer required, refill it to a level just below the secondary manway with condensate quality water treated with 50-150 ppm hydrazine (-15 liters of 35% hydrazine) and provide a nitrogen blanket overpressure. Then maintain chemistry control as directed in step 6.5.2.
NUCLEAR GENERATION SITE CHEMISTRY PROTEDURE S0123-III-2.1.1 UNITS 1, 2 AND 3 REVISION 2 PAGE 12 OF 16 6.0 PROCEDURE (Continued) 6.5.4 Condenser Hotwells (Dry Layup)
.1 The condenser hotwells should be layed-up with dry air (preferred) or nitrogen. The nitrogen could be supplied by hoses to the shell side of the reheaters or any other convenient point. Since the condensers are vented to the atmosphere via the turbine glands, it will not be possible to maintain any nitrogen overpressure. Various nitrogen purge rates may be tried.
.2 The condenser hotwell, when recommended by Chemistry Supervision, shall be opened, ventilated, and inspected for water. If water is present, it shall be vacuumed up. The source of water shall be found and an effort made to stop the in-leakage.
6.6 SALT LEAKS 6.6.1 Saltwater entry into the secondary system results in significant deterioration of the secondary system water quality. The initial indication of a salt leak is usually observed by a rapid increase in the hotwell quadrant cation conductivity and a cation conductivity alarm. If this occurs, -the following actions should be taken:
.1 Check the recorder values for cation conductivity for the four hotwell quadrants. Evaluate which quadrant appears to be the source of the in-leakage. Things to consider would be which quadrant ramped up first, which quadrant ramped the highest, etc. Request Operations to initiate their leak verification procedures for the suspect quadrant.
.2 Immediately collect and analyze all four quadrants for chloride. Use the chloride results to confirm that the proper quadrant was isolated and overboarded. When the samples are collected, check each quadrant to confirm the quality of the cation column and verify sample flow.
.3 Follow the sequence of events by observing the recorders for steam generator and condensate chlorides.
Use the bench conductivity bridge for continuous monitoring of the condensate system to determine peak value.
Record the value frequently, every 15 to 30 minutes. When the steam generator chlorides peak, take generator samples and run a complete set of analysis using step 6.3.2 parameters except activity.
NUCLEAR GENERATION SITE CHEMISTRY PROCEDURE S0123-III-2.1.1 UNITS 1, 2 AND 3 REVISION 2 PAGE 13 OF 16 6.0 PROCEDURE (Continued) 6.6 SALT LEAKS (Continued) 6.6.1.4 Check the generator chloride every four hours until the chloride is <500 ppb. At that time, reduce the sampling frequency to the applicable mode of operation.
The chloride results shall be graphed and filed in the Turbine lab for future evaluation.
.5 Perform all other analyses on the steam generators, condensate and feedwater at the frequency for the appropriate mode as detailed in thi's procedure.
6.6.2 If the salt leak is severe enough to shut the Unit down, a major turnover of water will be initiated.
Take the following actions:
.1 Condensate
.1.1 Request that the hotwells be drained to as low level as practical and refilled. Initiate a feed and bleed cleanup using condensate storage water. Check the cation conductivity and chloride every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Check all other parameters at the frequency detailed for the appropriate mode of operation.
.1.2 If instead of a feed and bleed cleanup for the condensate it is decided to drain and fill, check the system for pH, conductivity and chloride for the appropriate mode of operation within two hours after every refill.
.1.3 Once a shift, check each hotwell quadrant for chloride and cation conductivity. When the values match the condensate train, this special testing can be terminated.
.1.4 When the condensate chloride is <500 ppb, the sampling frequency should be reduced to the frequency required by the mode of operation.
.2.1 If the generators are on auxiliary feedwater, check every four hours for chloride. Check all other parameters at the frequency detailed in the appropriate mode of operation. Graph the chloride values and file for future reference.
NUCLEAR GENERATION SITE CHEMISTRY PROCEDURE SO123-III-2.1.1 UNITS 1, 2 AND 3 REVISION 2 PAGE 14 OF 16 6.0 PROCEDURE (Continued) 6.6 SALT LEAKS (Continued) 6.6.2.2.2 If the generators are being cleaned under the drain/fill process, check the generators for a complete set of Mode 1 parameters, except activity, within two hours after every refill.
.2.3 When the generators are <500 ppb, chloride sampling should be reduced to the frequency required by the mode of operation.
6.6.3 Chloride should be checked by titration when the value is expected to be >500 ppb. The Ion Chromatograph should be used when chloride is <500 ppb.
6.6.4 Following the initial cleanup, systems will be returned to service.
Every effort should be made to have these systems flushed prior to aligning them to the condensate train. As a precautionary measure, perform the following:
.1 Request Operations to inform Chemistry prior to initiating any system which could upset secondary water quality.
.2 When notified, observe cation conductivity whenever a condensate pump is started for the first time following a salt leak. Any significant increase should be brought to the attention of Chemistry Supervision for evaluation.
.3 Request Operations to isolate and initiate overboarding the appropriate hotwell quadrant prior to starting the circulating water pump. Observe the quadrant cation conductivity. If there is no significant continuing increase, request Operations to stop the overboarding.
.4 Observe the secondary system's chemistry parameter recorders when proceeding to Modes 1 and 2. If there is no significant increase noted, normal procedural sampling can be re-established.
6.7 DEFINITIONS D
= Daily M,W,F = Monday, Wednesday, Friday W
= Weekly 3/W = 3 times weekly M
= Monthly 2/W = 2 times weekly
NUCCEAR GENERATION-SITE CHEMISTRY PROCEIRE S0123-III-2.1.1 UNITS 1, 2 AND 3 REVISION 2 PAGE 15 OF 16 7.0 RECORDS 7.1 The following steam generator parameters shall be graphed:
- 1) pH
- 4) chloride
- 2) conductivity
- 5)
Na/PO, Molar Ratio
- 3) phosphate (Modes 1 and 2 only) 7.2 The following condensate and feedwater parameters shall be graphed:
7.2.1 Modes 1, 2, 3 and 4:
Condensate Feedwater
- 1) cation conductivity
- 1) pH
- 2) dissolved oxygen
- 2) Iron
- 3) chloride
- 3) copper 7.2.2 Modes 5 and 6:
Condensate and Feedwater
- 1) pH
- 3) chloride
- 2) conductivity 7.3 Graphs should be updated after each analysis is performed.
Notations should be made on the graphs to explain changes in parameters (for example, "blowdown" might be written to explain a decrease in steam generator conductivity). The Chemistry Foreman shall frequently review the graphs. As a minimum, the reviews shall be performed once each working day.
7.4 All data sheets will be reviewed and initialled daily (Monday-Friday) by the Chemistry Foreman.
7.5 Steam generator, condensate and feedwater data sheets shall be forwarded to COM with the Chemistry monthly letter.
7.6 Chart recorder paper shall be forwarded to the Shift Superintendent's office on an as-changed basis.
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rULJz K UtNtKAILN bllt LI1tM+/-ibKY tLJLtUUUi-4111-z.1.1 UNITS 1, 2 AND 3 REVISION 2 PAGE 16 OF 16 ATTACHMENT 1 IN-LINE MONITORS OF STEAM GENERATOR, CONDENSATE, AND FEEDWATER ALARM SETPOINTS PARAMETERS MONITORED DETECTOR NUMBER RECORDER NUMBER LOW HI HI-HI A,8,C S/G Conductivity CE 2,3,4 CR 3 125 A,B,C S/G pH pHE 4,5,6 pHR 1 9.4 10.2 AB,C S/G Activity RE 1216 RLR 1200.
A,B.C S/G Chloride ANE 3061 ANR 3061 500 A,B,C S/G Sodium ANE 3062 ANR 3062 20 30 Cation Conductivity Condenser Hotwells CE 17,18,19,20 CR 2 0.5 Condensate Cation Conductivity CE 7 CR 3 0.5 E,W Condensate Conductivity CE 8,10 CR 3 20 Condensate pH pHE 1 pHR 1 8.8 9.4 E,W Condensate Oxygen ANE 1-3,1-4 ANR 1 10 E,W Condensate Sodium ANE 3058 ANR 3058 80 100 E,W Condensate Chloride ANE 3057 ANR 3057 50 100 E,W Feedwater Conductivity CE 12,13 CRT 1 6.5 Feedwater pH pHE 3 pHR 1 8.8 9.4 E,W Feedwater Oxygen ANE 1-1,1-2 ANR 1 10 EW Feedwater Hydrazine ANE 2-2,2-1 ANRT 2 10 50 Main Steam Cation CE 1 CR 3 0.3 Conductivity A,B,C Steam Sodium ANE 3060 ANR 3060 45 60 A,BC, Steam Chloride ANE 3059 ANR 3059 10 20 0189c