ML20012A408
| ML20012A408 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 12/31/1989 |
| From: | FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML20012A405 | List: |
| References | |
| NUDOCS 9003090458 | |
| Download: ML20012A408 (9) | |
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CRYSTAL RIVER UNIT 3-ANNUAL REPORT OF FACILITY CHANGES, TEST, AND EXPERIMENTS IN' ACCORDANCE WITH 10 CFR 50.59(b).
JANUARY I - DECEMBER 31, 1989 t
In the attached report, each number refers to the Safety Evaluation questions listed below:
1..
Is the probability of occurrence or the consequences of an accident s
or malfunction evaluated in the FSAR. increased?
YES
, N0 I
2.
Is the possibility for an accident or malfunction of a different type than any evaluated. in the FSAR. created?
YES
, N0 3.
Is the margin of safety, as defined -in the bases for any Technical Specification reduced?
YES
, N0 7
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10 CFR 50.59 EVALUATIONS PERF0B!@
ON PLANT MODIFICATIONS IN ACCORDANCE WITH MAR PROCEDVRES FACILITY MODIFICATION-MAR 82-05-03-16' Closed' Circuit TV-(CCTV) Monitoring of Main Steam Relief Valves d
DESCRIPTION / SAFETY EVALUATION
-1.
No, this modification provides increased monitoring capability of the Main Steam Relief Valve (MSRV) releases. This CCTV system does not automatically actuate equipment or systems required to prevent or mitigate an, accident.
2.
No', the equipment serves a monitoring function only. Plant operation during normal and emergency-conditions is not affected by this modification.
3.
No, the addition of the CCTV system for the MSRVs will decrease the response time required for-the operators to - assess 'MSRV position and decrease dependence on dispatching of operators to remote locations.
FACILITY MODIFICATION MAR 89-03-06-01 Appendix R Chilled Water System DESCRIPTION / SAFETY EVALVATION 1.
No, the Appendix R Chilled Water (CH) System, and the new piping and valves added by this modification are classified non-safety related.
Also, th'is modification isolates the Seismic III portion of the CH System located in the Turbine Building from the Seismic I portions of the CH System by locking closed valves CHV-76 and CHV-77.
The portion isolated serves the Turbine Building switchgear room coolers which are non-safety related.
The load removed from the CH System (approximately 50 tons of refrigeration) is approximately 25% of the rated capacity of the CH System chillers (219 tons of refrigeration). The CH System chillers, prior to this modification, were loaded to 216 tons.
After completion of this modification, the CH Cystem load will' be approximately 75% of the capacity of the CH chillers.
The. compressors in the CH chillers, CHHE-1A and IB, can be adjusted to operate at loads as low as 40% of rated capacity. Therefore, the ability of the CH System to meet its safety functions is not adversely affected.
2.
.No, 'this modification isolates the Seismic III portion of the CH System (located in the Turbine Building) which served the non-safety related Turbine Building switchgear room coolers.
It also provides chilled water to these coolers from a non-safety related source, the Appendix R CH System.
The separation of the CH System and the Appendix R CH System required to meet Appendix R commitments is maintained:
the affected portions of the CH and Appendix R CH Systems and the new piping are located in a different i
fire area than the Seismic I portion of the CH System.
Also, the f
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- modification maintains the pressure boundary of the safety.related, Seismic ~
1 I portion _of the CH System to the requirements of the original design codes.
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- 3...No, this modification does not adversely affect the operability of the CH System to' provide chilled water -to components in - the Control. Complex.
Therefore, it does not adversely affect the control. room emergency-1 ventilation system nor does it affect the ability of any. other equipment.-
in the Control Complex to meet its design function.' Therefore, it does not
. reduce the margin of safety as defined in the. basis for. any Technical-Specification.
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g iv 10 CFR'50.59 EVALUATIONS PERFORMED ON FSAR TEXT CHANGES IN ACCORDANCE WITH h
PROCEDURE N0D-11 FSAR SECTION y_
'14.2.2.2 Steam Generat'or Tube Rupture Accident DESCRIPTION / SAFETY EVALUATION 1.-
No,Lthe drain lines being open (Main Steam Drain Traps; MSDTs) will allow-more radioactivity to pass into the condenser and ultimately into' the atmosphere, than originally assumed in the Steam Generator Tube Rupture-(SGTR) Accident. A reanalysis of the SGTR accident concluded that off-site doses would increase but still be less than-l% of 10 CFR 100 limits. This analysis ' also - assumed that operators continued to steam the affected 1
generator for up to eight hours, until the DH system could be activated, as opposed to the original analysis which assumed that the faulted 0TSG was isolated 34 minutes after the tube rupture.
This was done to assure that operator actions which could take place per EP-390 were considered in the safety analysis.
2.
No, allowing the drain lines to remain open will not degrade the. performance of, or increase the challenges to, existing safety systems assumed to function in-the accident analyses.
3.'
No, tM total off-site dose, with the MSDTs open during-a SGTR accident, w-ill be 'within 10 CFR.100 limits as shown by the B&W.SGTR Dose A-(B&WDocument 86-117463-01).
Continued steaming of the affected ge r..for up to eight hours during cooldown will also not result in
-exueoing 10 CFR 100' limits.
The initial conditions' assumed in the STGR analys.is and assured by the Limiting Conditions for Operation (LCOs)'
contained.in the Technical Specifications have not been revised.
FSAR SECTION 14.2.2.6 - Makeup System Letdown Line Accident DESCRIPTION / SAFETY EVALUATION
.1, No,- this revision to the CR3 FSAR Section 14.2.2.6 will change the location of the break assumed in the Makeup System Letdown Line Failure Accident.
The FSAR currently states that the assumed break is located. between the containment and the outboard isolation valve (MVV-49). The Analysis Basis Document (ABD) Analysiscis based upon a break which is located downstream of the outboard isolation valve (MUV-49).
B&W has reviewed the accident analysis which was performed and has confirmed the accuracy of the ABD.
No physical changes to the plant are required as a result of this change.
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2.
No, physical changes to the plant are not required es a result of this change.
Thk revision to the CR3 FSAR Section 14.2.2.6 will change the
. location of the break assumed in the Makeup System Letdown Line Failure Accident.
The FSAR currently states that the assumed break is located i
between the containment anc' the outboard isolation valve (MVV 49).
The
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Analysis Basis Document (ABD) Analysis is based upon a break which is located downstream of the outboard isolation valve (MVV-49).
B&W has i
reviewed the accident analysis and has confirmed the accuracy of the ABD.
l 3.-
No, physical enanges to the plant are not required as a result of this change.
This revision to the CR3 FSAR Section 14.2.2.6 will change the location of the break assumed in the Makeup System Letdown Line Failure Accident.
The FSAR currently states that the assumed break is located between the. containment and the outboard isolation valve (MUV-49).
The Analysis Basis Document. (ABD) Analysis is based upon a break which is
-located downstream of the outboard isolation valve (MUV-49).
B&W has reviewed the actual accident analysis and has confirmed the accuracy of the ABD.
The margin of. safety, as defined in the basis for any Technical Specification is not reduced.
FSAR SECTION 14.2.2.6 - Makeup System letdown Line Accident DESCRIPTION / SAFETY EVAll'ATION 1.
No, the present FSAR analysis for the Make-Up System letdown line rupture takes credit for 90% filtration of iodines by the Auxiliary Ruilding Ventilation System (ABVS) prior to release of the activity to the environment. The analysis also assumes a loss of offsite power coincident with the line break. Since the ABVS is a'non-safety-related system and is not aligned to received emergency power from the diesel generators, it will not be available for mitigation of this postulated accident.
An analysis was performed to determine the offsite doses due to this postulated accident without credit for iodine filtration by the ABVS. The results showed an increase in the offsite thyroid doses by a factor of ten, which is in proportion with the decrease in assumed ABVS charcoal filter efficiency (90% vs. 0%).
The whole body doses increased by 1.0 millirem at the EAB and 0.1 millirem at the LPZ. These are increases of 1.5 percent and 1.7 percent, respectively.
These revised doses are much less than the limits specified by 10 CFR 100.
The 1976 SER for CR3 does not address this accident in the list of' events reviewed by the staff.
Therefore,10 CFR 100 is consioered to be the acceptance limit for protection of the public and no unreviewed safety question within the meaning of 10 CFR 50.59 exists.
Since the increased doses are still within the original acceptance limit that is the licensing L
basis for CR3, the answer to this question is "no".
This philosophy is consistent with NRC comments on NSAC-125 " Guidelines for 10 CFR 50.59 Safety gf M 1uations".
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2.
No, this change pertains only to the operation of the ABVS during a l
postulated Make l? System letdown line rupture design basis accident as presently evaluated in the FSAR.
Neither the design nor operation of the i
r ABVS has been altered. Therefore, no new accident or malfunction will be created by this change.
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3.
No, the increased offsite doses are still less than one percent of the safety limits specified by 10 CFR 100.11. Therefore, the margin of safety i
as defined in the technical specifications has not been decreased.
FSAR SECTION J
Table 411 - Steam Generator Feedwater Quality Specifications
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DESCRIPTION / SAFETY EVALUATION 1.
No, the proposed change for pH limits in the secondary cooling loop at CR-l 3 does not increase the probability of an occurrence of an accident or malfunction of equipment important to safety as evaluated in the FSAR.
Since this change in pH limits affects only the secondary side, only malfunctions or accident scenarios involving secondary side piping or components need to be considered.
The only possible adverse consequence of the increase in pH would be increased corrosion of copper alloy components (i.e., the condenser) which could eventually result in leaks, i.e., loss of integrity of the secondary loop. However, in order for the probability of an accident or malfunction involving loss of secondary loop integrity to be increased, the increased secondary side pH would have to cause a significant increase in condenser tube corrosion over that currently occurring. Experience and test programs at Crystal River-3, however, have i
shown that o increased pH should not significantly increase the corrosion of the conds.r tubes. Further, chemistry program procedures are in place to limit the concentration of dissolved copper in the feedwater, thus, t
effectively limiting the corrosion of the condenser to acceptable levels.
Morpholine is currently used for pH control at CR-3.
French [1] and Canadian [2] experience suggest that morpholine is less aggressive toward copper alloys (e.g., condenser tubes) than ammonia.
The increased pH will result in a decrease in the corrosion of iron alloy i
components and piping and, therefore, will increase the overall long-term integrity of that portion of the secondary plant.
Lower iron alloy i
corrosion-rates will result in an overall decrease in corrosion product concentrations in the feedwater. Lower concentration of corrosion products in the feedwater results in decreased fouling of the steam generators.
Inasmuch as steam generator deposits can provide a mechanism for concentrating corrosive feedwater impurities next to the tube surface l
(although deposit analyses and steam generator experience to date indicates I
that.this is not occurring to a measurable degree at CR-3), reducing steam l
generator fouling also minimizes the potential for the initiation of at l
1 east some types of steam generator corrosion. Industry experience and test programs have shown that increasing the pH will not adversely impact steam generator tube integrity.
Increasing the secondary plant pH will, therefore, not have a negative impact on steam generator tube or secondary cycle piping integrity.
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The increased pH in the secondary cycle at CR-3 will not increase the 5
consequence of an accident or malfunction of equipment important to safety L
as evaluated in the FSAR. Even if it is assumed that the increased pH would increase the corrosion of the condenser, the consequences of accidents or i
malfunctions in the FSAR would not be increased since those accident
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scenarios are essentially worst case losses of secondary cycle integrity, e.g., double-ended ruptures of a steam generator tube (Section 14.2.2.2),
J main feedwater line (Section 14.2.2.9), or a steant line (Section 14.2.2.1).
The production of such double ended ruptures of such iron alloy components simply by increased corrosion would be highly unnatural and not reasonably expected by operation with an increased secondary plant pH.
A loss of electric power concurrent with a 1 gpm steam generator tube leak was also evaluated in the FSAR (Section 14.1.2.8).
The consequences of this malfunction would not be increased even if the increased pH did result in' more corrosion than that observed under the current pH 11' nits since Plant Technical Specifications (Section 3.4.6.2) prevent continued plant operation -
with a SG tube leak of I gpm or more.
REFERENCES FOR PART 1 i
1.
Proceedings of the EPRI Meeting on Advanced Water Chemistry i
Qualification Requirements, May 28 29, 1987 2.
" Survey of Domestic and Foreign PWR Experience with Morpholine in Chemistry Control by All-Volatile Treatment", EPRI Final Report NP-467), July 1986 t
2.
No, an increase in pH levels in the secondary cooling loop at CR 3 does not create the possibility for accidents or malfunctions of a different type other than those evaluated in the FSAR. If the increased pH levels results I
in increased corrosion of the copper alloy condenser tubes, then the -
possibility of loss of secondary system integrity would be created.
Corrosion would produce leaks in secondary system outside the radiation controlarea(RCA). Assuming that a primary-to secondary leak is occurring in the steam generators concurrent with condenser leakage, this could result in radioactivity being released to the environment via pathways outside the RCA, i.e., through the turbine building and via the condenser vacuum pumps.
Since a vacuum is maintained on the shell side of the condenser during operation, any condenser leaks concurrent with a steam generator tube leak will result in salt water being drawn-in through the tube defect, and not radioactive steam escaping out of the condenser via the tube leak. Activity releases via the turbine building and condenser vacuum pumps are evaluated in Section 11.2.3.3 and Section 14 of the FSAR.
Accident or malfunction scenarios involving steam generator tube leaks and/or secondary system piping or component leaks (that challenge the integrity of the fuel or the reactor coolant system) are evaluated in Section 14 of the FSAR (see l
especially Sections 14.1.2.8, Loss of Electric Power; 14.2.2.1, Steam Line 1
Failure; 14.2.2.2, Steam Generator Tube Failure; and 14.2.2.9, Main l
Feedwater Line Break.
3.
No, an increase in secondary plant pH pertains to only three Technical Specification 3/4.4.5, 3/4.4.6.2, and 3/4.7.1.6.
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Section 3.7.1.6 does not specify any limits for secondary water chemistry.
l Therefore, increasing the pH does not affect this Technical Specification.
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r Section 3.4.6.2 limits primary to secondary leakage through steam generator tubes to a total of I gpm.
The margin of safety defined in the basis of 3
this Technical Specification (i.e.,
ensuring that the radiation dose contribution from SG tube leakage will be limited to a small fraction of Part 100 limits in the event of either a SG tube rupture or steam line
- t break) is not reduced since, as noted in the previous response to Question 1, the increase in secondary plant pH does not increase the potential for steam generator tube corrosion over that currently present with existing pH limits.
The basis for Technical Specification 3/4.4.5 states that "the plant is s
expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits fond to result in negligible A
corrosion of the steam generator tubes."
As noted in 1 above, plant i
experience and test programs have demonstratec that increased pH levels does result in negligible corrosion of the steam guerator tubes {1.e., the potential for steam generator tube corrosion will os no greater than that currently present under existing pH limits).
Further, to the extent that secondary cycle piping corrosion is reduced by the increased pH levels, the potential for the initiation of under depesit corrosion phinomena of the steam generator tubes is lessened as long us coppr corrosion is not increased. Plant procedures (e.g., CP-138) are in effbet to restrict copper corrosion to acceptable levels as defined by industry standards (e.g., EPRI, INPO).
Therefore, the margin of safety as definnd in the basis for this Technical Specification is not reduced.
FSAR SECTION Table 2-1 Chemical Storage DESCRIPTION / SAFETY EVALVAT103 1.
No, the valves presented in this table were reviewed by the NRC as part of the Control Room Habitability Study. The NRC issued SER accepting the study on May 25, 1989.
2.
No, the accidents and malfunctions considered due to chemical storage have been evaluated and approved by the NRC for the CR-3 Control Room 2
Habitability Study.
3.
No, the Technical Specification limits for chemicals entering the control room are based upon the values presented in these tables.
No margin of 4
safety is reduced.
FSAR SECT 10ti 1.7 Quality Program (Operational)
DESCRIPTION / SAFETY EVALUATION l.
No, this change affects the Nuclear Operations Organizational structure,
.but not the functions and, therefore, does not increase the probability of occurrence or increase the consequences of an accident or malfunction of i
equipment previously evaluated in the FSAR.
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2.
No, this administrative change does not affect the FSAR by creating a L
different type of accident or malfunction of a different type than any previously evaluated in the FSAR.
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3.
No, this administrative change does not affect the Tech Specs.
FSAR SECTION 14.2.2.5 - Loss of Coolant Accident DESCRIPTION / SAFETY EVALUATION i
1.
No, this change adds a clarification to the LOCA section to indicate NRC approval of the SBLOCA model, i
2.
No, the description added to the FSAR reflects NRC approval of the SBLOCA model and its acceptable use for CR-3.
r 3.
No, the change describes the NRC correspondence which approves the SBLOCA model.
Margin of safety has been considered by the NRC when granting
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approval.
FSAR SECTION
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3.2, 9.7, 10.5,
14.2 DESCRIPTION
/ SAFETY EVALUATION 1.
No, these changes are editorial.
Corrections were made in spelling, punctuation, sentence structure, and/or changes in table labels.
2.
No, modifications are not made by these changes.
Spelling, punctuation, etc. can not create a different type of accident.
3.
No, these changes have not affected any Technical Specification and, therefore, the margin of safety is not reduced.
FSAR SECTION 14.2.2.8 - Waste Gas Decay Tank (WGDT) Rupture Accident DESCRIPTION / SAFETY EVALVATION 1.
No, this change makes clarifications regarding assumptions in the WGDT Rupture Accident analysis which are presently reflected incorrectly in the i
FSAR. There is no change to the dose consequences of this event.
l 2.
No, this change corrects the description of an analysis for an accident presently in the FSAR. No change to plant operation or design is involved.
i ll Therefore, no new accident or malfunction is created.
3.
No, no change to the WGDT Rupture Accident offsite doses occur as a result of this FSAR text change. Therefore, the safety limit of 10 CFR 100 is not exceeded and the margin of safety remains the same.
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