ML18018A599
| ML18018A599 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 07/05/1983 |
| From: | Mcduffie M CAROLINA POWER & LIGHT CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| LAP-83-110, NUDOCS 8307080315 | |
| Download: ML18018A599 (16) | |
Text
REGULATOR NFORMATION DISTRIBUTION S EM (RIDS)
AC'CESS/ONNBR!8307080315 OOG.DATE: 83/07/05 NOTARIZED!, NO FACIL:50.400 Shearon Har,.rjs NuCl'ear Power Planti Unjt ii ICaroljna
. 50 401 Shearon Harris Nuclear Power'lanti Unit 2i 'Carolina AUTH NAME,,
AUTHgR AFFILIATION MCDUFF/EsM A ~
Car ol ina Power L Liaht Co RECIP.N/ME RECIPIENT 'AFFILIATION.,
DENTONSH R.
Of f ice'f Nuc 1 ear Reactor Reaul at i one Di r ector'OCKET 0
05000400 05000401'SUBJECTs Forwards responseto graft 'SER Open Item 209 failures assumed"-for accidents'ISTRIBU]'ION CODE:
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~TITLE: Licensina Submittal:
PSAR/FSAR Amdts 8 Related NOTES' r e sinai e T~s r
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Carolina Power & Light Company JUL OS >9S3 SERIAL:
LAP-83-110 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear. Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NOSo 1
AND 2 DOCKET NOS. 50-400 AND 50-401 DRAFT SAFETY EVALUATION REPORT RESPONSES REACTOR SYSTEMS BRANCH
Dear Mr. Denton:
Carolina Power.
& Light Company (CP&L) hereby transmits one original and forty copies of the response to the Shearon Harris Nuclear Power Plant Draft Safety Evaluation Report (DSER)
CP&L Open Item 209.
Carolina Power
& Light Company will be providing responses to other DSER Open Items shortly.
Yours very truly, PS/ccc (6659PSA)
Attachment M. A. McDuffie Senior Vice President Engineering
& Construction CC:
Mr.. N. Prasad Kadambi (NRC)
Mr. Evangelo Marinos (NRC-RSB)
Mr.
G. F. Maxwell (NRC-SHNPP)
Mr. J.
P. O'Reilly (NRC-RII)
Mr.. Travis Payne (KUDZU)
Mr. Daniel F.
Read (CHANGE/ELP)
Chapel Hill Public Library Wake County Public Library Mr. Wells Eddleman Dr. Phyllis Lotchin Mr. John D. Runkle Dr. Richard D. Wilson Mr. G. 0. Bright (ASLB)
Dr. J.
H. Carpenter (ASLB)
Mr. J. L. Kelley (ASLB) 83070803i5 830705 PDR ADOCK 05000400 E
PDR 411 Fayetteville Street
~ P. O. Box 1551 o Raleigh, N. C. 27602
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Shearon Harris Nuclear Power Plant (SHNPP)
Draft Safety Evaluation Report (DSER)
Reactor Systems Branch 0 en Item 209 (DSER Section 15.0 pa e 15-1)
The staff has asked the applicant to supply a listing of assumed single failures utilized in the Final Safety Analysis Report (FSAR) Chapter 15 analysis and the limiting single failure in each event analyzed that results in the peak pressure or limiting fuel performance for each event.
The applicant should discuss, in the analysis of each event, the way that the limiting single failure is selected.
Response
All of the transients analyzed in Chapter 15 were analyzed assuming the most limiting single failure (e.g., loss of one protection signal of safety injection (SI) train failure).
Table I lists the limiting'ingle failures for each American Nuclear Society (ANS) Condition II event (faults of moderate frequency).
Table II lists the limiting single failures for Non-Condition II events.
The single failures listed in Tables I and II are the limiting failures for their respective event.
Operator error is not explictly considered in accidents analyzed in Chapter 15.
I.
Sin le Failures Assumed for Accidents of 1foderate Fre uenc The incidents of moderate frequency were analyzed consistent with the acceptance criteria given in the Standard Review Plan (SRP) concerning peak pressure (less than 110 percent of design), fuel integrity (departure from nucleate boiling ratio (DNBR) limit), generation of more serious plant conditions, and single active failures.
Pressure transients for each event are provided in the FSAR and demonstrate that the pressure remains below 110 percent of design pressure.
Fuel cladding integrity is demonstrated for each case by showing that the DNBR remains above the limit value.
This is discussed in the results and conclusions sections for each event.
For each transient, its associated worst single failure within the protection system assumed in the FSAR analyses is given in Table I.
The protection system is defined as those safety functions required to mitigate the consequences of the event.
This includes not only the Solid State protection System (SSPS),
but also the Engineered Safeguards Features (ESF) and pressurizer and steam generator safety valves.
Res onse to 0 en Item 09 (Continued)
These single failures were selected based on the requirements of 10CFR50 Appendix A, the
- SRP, and Reg.
Guide 1.53 (which addresses IEEE-279 and IEEE-379).
A single failure is ".
. an occurrence which results in the loss of capability of a component to perform its intended safety functions.>>
The single failure criterion states that a "single failure within the protection system shall not prevent proper protective action at the system level when required" (IEEE-279).
The single failures which are considered are active failures, consistent with the SRP acceptance criteria.
Failures in the protection system which are not required to mitigate the consequences of an accident are not considered.
These're failures of systems which are not challenged during the transient and are not active failures.
Such failures are independent failures and are therefore not within the scope of the evaluation.
For each event listed in Table I, a brief discussion of the assumed single failure is provided in this discussion.
The purpose of the discussions is to justify that the single failure assumed is indeed the worst single failure.
These failures are failures at the system level and consider the failure of a protective function.
The cause or mechancial nature of the failure which causes the system failure is not discussed, since these are addressed in the failure modes and effects analyses (FMEA's) of the SSPS and ESF and in Chapters 6, 7, and 9 of the FSAR.
Therefore, further detail beyond the systems level single failure of loss of one protection train is not provided.
The steam generator safety valves may be required to prevent a
pressurization of the secondary system.
Except where it is already stated in the FSAR, the steam generator valves are not challenged or required to mitigate the consequences of the event.
Failures of these valves are not considered since they are not active failures.
These independent failures are not applicable.
Therefore, failure of these valves is not discussed below unless they are actuated as stated in the FSAR.
Finally, a loss of offsite power is not considered as a single failure for these events.
The SRP does not require consideration of a loss of offsite power for the accidents listed in Table I (loss of AC power, FSAR Section 15.2.6, is by definition an exception).
Furthermore, no single active failure will cause a loss of offsite power to the emergency buses.'herefore, consideration of this failure is not applicable.
Feedwater Tem erature Reduction (FSAR Section 15;1')
As stated in FSAR Section 15.1.1.1, this event is similar to the effect of increasing steam flow.
This is bounded by the event in FSAR Sections
- 15. 1.2 and
- 15. 1.3, as stated in FSAR Section
- 15. 1. 1.3.
Res onse to 0 en Item 209 (Continued)
Excessive Feedwater Flow (FSAR Section 15.1.2)
As seen in FSAR Figure 15.1.2-1, the pressurizer pressure eventually decreases until the time of turbine trip.
The pressure rise is caused by the conservative delay between turbine trip and reactor trip; however, the pressurizer power operated relief valves (PORV's) and safety valves do not open.
Since they are not required to mitigate the consequences of the event, a single failure in these valves is not applicable and has no impact.
Failure of a feedwater isolation valve (FWIV) to close will have no impact since the DNBR is already increasing by the time the FMIV closes (FSAR Table 15.1.2-1).
The engineered safeguards features are not required for this event.
Therefore, a single failure in the ESF is not applicable and has no impact.
Therefore, the failure of one protection train as listed in Table I is the limiting single active failure.
Excessive Steam Flow (FSAR Section 15.1.3)
As stated in FSAR Section 15.1.3.2, the plant reaches a stabilized condition.
No reactor trip is required, no pressurizer relief valves are required to reduce pressure (FSAR Figures 15.1.3-1, 15.1.3-3, 15.1.3-5, 15.1.3-7),
and no ESF actuation occurs.
Since the protection system is not required to function for this event, a single failure does not apply and has no impact.
Inadvertent Secondar Depressurization (FSAR 15.1.4)
As stated in FSAR Section 15.1.4.1, it is the failure (opening) of a steam dump, relief, or safety valve which initiates the transient.
As seen in FSAR Figure 15.1.4-3, this is a depressurization event, therefore, pressure relieving functions of the protection system are not challenged nor required to mitigate the consequences of the event.
The only portion of the protection system required is the safety injection portion of the ESF.
A single failure in a protection train of the signals which actuate SI (FSAR Section 15.1.4.1 Item a) will have no impact due to the redundancy, diversity, and independence of the SI actuations signals.
The failure of one SI train (listed in Table I) is the limiting single failure since it reduces SI flow, delays the injection of boron to the core,
- and, consequently allows a "closer" return to criticality.
This is the single failure assumed in the FSAR as stated in FSAR Section 15.1.4. 2.
For this event, the DNB design basis is met by demonstrating no return to criticality (FSAR Section 15.1.4.3).
Loss of External Load (FSAR Section
- 15. 2 ~ 2)
This is bounded by the event described in FSAR Section 15.2.3, as stated in FSAR Sections 15.2.2.1 and 15.2.3.1.
Res onse to 0 en Item 209 (Continued)
Unlike a depressurization transient, for this analysis, the ability to maintain Reactor Coolant System (RCS) pressure below 110 percent of design per the SRP criterion must be explicity addressed.
Since the DNBR increases with pressure (assuming all other variables are held constant),
the event is analyzed with and without pressure. control to address both peak pressure and DNBR concerns.
As stated in FSAR Section 15.2.3.2, both the pressurizer.
and steam generator safety valves may be required to operate.
Assumptions relative to their operation are described under Items 4 and 5 in the FSAR.
If the pressurizer relief/safety valves fail to close once the pressure has been reduced, there will be no impact on the minimum DNBR.
This is because the valves are not required to close until after the time of reactor trip, at which point the DNBR is rising and is very high (see FSAR Figures 15.2-1 through 15.2-8).
As stated in FSAR Section 15.2.3. 2, Item 4, steam relief is obtained by the steam generator safety valves.
- However, these or any other steam relief valves would not be required to close until after reactor trip, when both the RCS pressure and DNBR are past their.
maximum and minimum values respectively.
Therefore, failure to close would have no impact.
Although the ESF may be required to function to supply auxiliary feedwater, a failure in the ESF would have no impact since credit for.
auxiliary feedwater is not taken (FSAR Section
- 15. 2.3', Item 6) ~
Therefore, the limiting single failure is one protection train (Table I).
Inadvertent Closure of Main Steam Isolation Valve (MSIV) (FSAR Section 15.2.4)
This is bounded by FSAR Section 15.2.3 as stated in the FSAR.
Loss of Condenser Vacuum (FSAR Section 15.2.5)
The results and conclusions of FSAR Section 15.2.3 apply to this event as stated in the FSAR.
Loss of AC Power (FSAR Section 15.2.6)
For this event, the ability of the protection system to provide long term cooling is verified.
The loss of one auxiliary feedwater pump of the ESF is the limiting single failure, as stated in Table I.
A reduction of auxiliary feedwater capacity reduces the capability of the auxiliary feedwater to provide long term cooling.
This results in a higher primary side heatup and pressure.
The pressure transient of FSAR Figure
- 15. 2.6-3 shows that the pressurizer safety valves are actuated for this event.
Failure of the valves to close would have no impact since the auxiliary feedwater is adequately removing the decay heat by that time (Table I).
For the case where the single active failure is the failure of the pressurizer PORV or safety valve to close, credit can be taken for complete auxiliary feedwater capability.
This would reduce the peak pressure and cause the time at which decay heat equals
Res onse to 0 en Item 209 (Continued) heat removal capability to be sooner.
As stated in FSAR Section
- 15. 2.6.1, the steam generator safety and relief valves are used to dissipate decay heat during long term cooling.
Since it is desirable to have these valves
- open, failure to close has no impact, especially since the emergency feedwater supplies sufficient heat removal capability.
Single failures which result in loss of signals which actuate auxiliary feedwater, reactor trip, or valve openings have no impact due to their redundancy, diversity and independence.
Therefore, the single failure listed in Table I is the loss of one auxiliary feedwater pump.
Loss of Normal Feedwater (FSAR Section
- 15. 2.7)
As for the loss of power event, the primary concern for the loss of normal feedwater is long term cooling capability which is provided by the emergency feedwater system.
Therefore, as for the loss of AC power, the single active failure causing the loss of one auxiliary feedwater pump is the limiting single failure, as stated in FSAR Section 15.2.7. 2.
Loss of Flow (FSAR Sections 15.3.1 and 2)
The protection for this event is discussed in FSAR Sections 15.3.1. 2 and 15.3. 2. 2.
A single failure in the ESF is not applicable since the ESF are not required to mitigate the consequences of the event.
As can be seen in FSAR Figures 15.3.1-2 and 15.3.2-2, the pressurizer PORV's may open.
- However, failure to close will have no impact since the point of minimum DNBR is past and the DNBR is rising by the time the valves close (FSAR Figures 15.3.1-4 and 15.3. 2-4).
Therefore, the worst single failure is that of one protection
- train, as stated in Table I.
Rod Cluster Control Assembl (RCCA) Bank Mithdrawal from Subcritical (FSAR Section 15.4.1)
Although the pressure transient is not shown for this transient, an increase in RCS pressure is expected due to the increase in heat flux and temperature.
However, if the PORV's opened and failed to close, there would be no impact on the minimum DNBR since credit for the change (increase) in pressure is not taken in the DNBR analysis.
The ESF are not required for this
- accident, therefore, a single failure in the ESF is not applicable.
Therefore, a loss of one protection train is the limiting single failure.
RCCA Bank Withdrawal at Power (FSAR Section 15.4.2)
This event is primarily a DNB event and demonstrates the adequacy of the over-temperature AT and high flux trips, as stated in FSAR Section 15.4. 2.3.
Typical transients for the RCCA bank withdrawal at power event are provided in FSAR Figures 15.4.2-1 through 15.4.2-9.
Operation of pressure relieving valves would serve to reduce pressure and thus minimize the DNBR.
0 wa
Res onse to 0 en Item 209 (Continued)
(If no pressure control was available, the maximum pressure would be limited to that which results in a high pressurizer pressure trip.
This is a less limiting pressure transient than those events discussed in FSAR Section 15.2.)
Failure of valves to close would have no impact, since the point of minimum DNBR is past by t'e time the pressure begins to fall (after trip) as seen in the transient figures.
As discussed in FSAR Section 15.4. 2.1, for some
- cases, the steam generator safety valves are opened.
However, failure to close has no impact since the point of minimum DNBR comes right after reactor trip.
Failures in the ESF are not applicable since the ESF are not required.
Therefore, the worst single failure is one protection train as stated in Table I.
The worst single failure for this event is the failure of one Nuclear Instrumentation System (NIS) channel.
This results in fewer dropped RCCA's being detected in order to initiate reactor trip via negative flux
- rate, but has no impact if no trip is generated (i.e., if credit for trip is not taken because of the failure).
The plant reaches a new equilibrium condition, and no further protective action is required.
Therefore, consideration of other single failures within the protection system is not applicable.
Staticall Misali ned RCCA (FSAR Section 15.4.3)
As stated in Table I, no transient analysis is required.
Furthermore, no protective functions are required and single failures have no
.impact.
Inactive Reactor Coolant Pump Startu (FSAR Section 15.4.4)
The pressure transient in FSAR Figure 15.4.4-4 shows that the pressurizer PORV's may be challenged for this event.
However, failure to close would have no impact, since the point of minimum DNBR is past by the time the failure could occur (FSAR Figure 15.4.4-5).
Failures in the ESF are not applicable since the ESF is not required to mitigate the consequences of the event.
Therefore, the limiting single failure is the failure of one protection train, as stated in Table I.
Inadvertent Actuation of the ECCS (FSAR Section 15.5.1)
As stated in FSAR Section 15.5.1.1, it is a failure in the ESF which initiates the event.
As seen in FSAR Figure 15.5.1-2, this is initially a depressurization event.
Res onse to 0 en Item 209 (Continued)
The pressure then rises to the PORV setpoint.
The PORV's are capable of maintaining system pressure below 110 percent of design.
Failure of the PORV's to close would have no impact on the DNBR, since it is already high and never falls below the initial value (FSAR Figure 15.5.1-3).
Therefore, the failure of one protection train listed in Table I is the limiting single failure.
Increase in RCS Inventor (FSAR Section 15.5. 2) 15.4.6.
As stated in the PSAR, this is bounded by FSAR Sections 15.5.1 and Inadvertent RCS De ressurization (PSAR Section 15.6.1)
As stated in PSAR Section 15.6.1.1, it is a single failure resulting in the opening of a pressurizer PORV or safety valve which initiates the transient.
Although ESF features might be actuated, they are not required to mitigate the consequences of the event, since the DNBR rises after reactor trip.
Therefore ESP failures are not applicable.
Therefore, the worst single failure is failure of one protection train.
Failure of Small Lines (FSAR Section 15.6.2)
No transient analysis is involved for this event.
The protective system is not required to function, since operator action terminates this event as stated in FSAR Section 15.6. 2.
II.
Sin le Failures for. Non-Condition II Events Table II addresses events other than those of moderate frequency.
Fuel-failures (if any) which occur are discussed in the appropriate section of the FSAR.
Although operator error is not explicitly considered in the accidents analyzed in Chapter 15, it should be noted that cognitive operator errors may induce the transient or cause the limiting single failures listed.
With the exception of the feedline break (PLB), the steamline break (SLB), the steam generator tube rupture (SGTR),
and the boron dilution, no credit for operator action or non-action is considered during the transient.
For the boron dilution event, operator action to terminate the dilution is taken within the time frames specified in the FSAR.
Por the SLB, FLB, and SGTR events, operator action (subsequent to the action of protection system) is considered to mitigate the consequences.
Action is assumed to be complete within 10 minutes, 30 minutes, and 30 minutes respectively.
Response
to 0 en Item 209 (Continued)
The impact of the operator's failure to take action specifically for the SLB, FLB, and SGTR are discussed in the CP&L's response to the Safety Review +estion 440.65, submitted to the Staff on August 2,
1982.
TABLE I SINGLE FAILURES ASS@AD FOR ACCIDENTS OF MODERATE FREQUENCY Event Description Section Worst Failure Assumed Effect Feedwater temperature reduction Excessive feedwater flow Excessive steam flow 15 ~ 1 ~ 1 15 ~ 1. 2 15.1.3 One protection train (1) none none none Inadvertent secondary depressurization 15.1.4 One Safety injection train delays boron Loss of external load Turbine trip Inadvertent closure of MSIV Loss of condenser vacuum Loss of ac power Loss of normal feedwatez
- 15. 2. 2
- 15. 2.3
- 15. 2.4
- 15. 2.5 15 ~ 2.6
- 15. 2.7 One protection train One protection train One protection train One protection train to core none none none none One auxiliary feedwater pump increases primary heatup One auxiliary feedwater pump increases primary heatup none none none Loss of forced reactor coolant flow 15.3.1&2 One protection train One protection train One protection train 15.4.2 15.4.3 RCCA bank withdrawal at power Dropped RCCA, dropped RCCA bank One nuclear instrumentation none RCCA bank withdrawal from subcritical 15.4.1 Statically misaligned RCCA Single RCCA withdrawal Inactive RC pump startup 15.4.3 15.4.3 15.4.4 system channel (2)
One protection train Standby charging pump is operating none none Reduces time to criticality
TABLE I (Continued)
SINGLE FAILURES ASSUMED FOR ACCIDENT OF MODERATE FREQUENCY Event Descri tion Section Worst Failure Assumed Effect Uncontrolled boron dilution Inadvertent ECCS operation at power Increase in RCS inventory Inadvertent RCS depressurization Failure of small lines carrying primary coolant outside containment 15.4 '
15.5.1 15.5.2 15.6.1 15.6. 2 Standby charging pump is operating One protecti'on train One protection train One protection train (2)
Reduces time to criticality none none none none (1)
No protective action required.
(2)
No transient analysis involved.
10
TABLE II SINGLE FAILURES FOR NON-CONDITION II EVENTS Event Description Section Worst Failure Assumed Effect Single Rod Withdrawal Inadvertent Fuel Loading LOCA (Small Break)
Gaseous Rad Waste Failure Liquid Rad Waste Failure Liquid Tank Failure Fuel Cask Drop Steamline Rupture Feedline Rupture Locked Rotor Shaft Break Rod Ejection Steam Generator Tube Rupture Fuel Handling Accident 15 4.3 15.4.7 15.6 5
15.7.1 15.7.2 15.7.3 15.7.5 15.1.5
- 15. 2.8 15.3.3 15.3.4 15.4.8 15.6.5 15.7.4 One Protection Train (1) (2)
Loss of One SI Train (1)
(1)
(1)
(1)
Loss of One SI Train Loss of One Aux Feed Pump None None Higher PCT None None None None Delay Boron to Core Delay Cooling Minimum Aux Feed (e.g.,
One Pump)
(1)
Increased Steam Release None Loss of One Protection Train None Loss of One Protection Train None Loss of One Protection Train None (1)
No Protective Action Required.
(2)
No Transient Analysis Involved.
(6659PSAccc)