ML20210C290
| ML20210C290 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 02/03/1987 |
| From: | Schnell D UNION ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| ULNRC-1442, NUDOCS 8702090360 | |
| Download: ML20210C290 (30) | |
Text
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i Union
.Etscraic PE 1901 Gratiot Street St. Louis Donald F. Schnell Vce President February 3,
1987 U. S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.C.
20555 Gentlemen:
ULNRC-14 4 2 DOCKET NUMBER 50-483 CALLAWAY PLANT STEAM GENERATOR TUBE RUPTURE ANALYSIS
References:
- 1) Letter from P.W. O'Connor to D.F. Schnell dated 11/12/86
- 2) ULNRC-1424 dated 12/30/86
- 3) SLNRC-86-01 dated 1/8/86 Enclosed is the response to Enclosure 2 of Reference 1 which requested additional information regarding the Steam Generator Tube Rupture Analysis for the Callaway and Wolf Creek Plants.
This response is being submitted in accordance with the schedule established in Reference 2.
The enclosure to this letter is identical to that being submitted by Wolf Creek Nuclear Operating Corporation in re'sponse to a similar request.
It is expected that the review of this response will continue to be a single review for both Callaway and Wolf Creek, in the same manner as the review of the original Steam Generator Tube Rupture Analysis Report submitted via Reference 3.
If you have any questions concerning this matter, please contact us.
Very trul ou s, Donald F.
Schnell DS/mc Enclosures P
A00l
$l Maihng Address: P.O. Box 149, St. Louss MO 63166
STATE OF MISSOURI )
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^
CITY OF ST. LOUIS )
Donald F.
Schnell, of lawful age, being first duly sworn upon oath says that he is Vice President-Nuclear and an officer of Union Electric Company; that he has read the foregoing document and knows the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information.and belief.
By 6onald F. S'chnell Vice President Nuclear 8 N day of [s & go o_. 198%
SUBSCRIBED and sworn to before me this f
Or/
BARBARA b FAFF NOTAPY PUBLIC, STATE OF MISSOURI MY COMM;SSION EXPIRES APRIL 22, 1989 ST. LOUIS COUNT.Y.
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cc:
Gerald Charnoff, Esq.
Shaw, Pittman, Potts & Trowbridge 1800 M. Street, N.W.
Washington, D.C.
20036 Nicholas A.
Petrick Executive Director SNUPPS 5 Choke Cherry Road Rockville, Maryland 20850 W.
L.
Forney Division of Projects and Resident Programs, Chief, Section lA U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 Bruce Little Callaway Resident Office U.S. Nuclear Regulatory Commission RR$1 Steedman, Missouri 65077 l
Paul O'Connor (2)
Office of Nuclear Reactor Regulation U.S.- Nuclear Regulatory Commission Mail Stop 316 7920 Norfolk Avenue Bethesda, MD 20014 Manager, Electric Department Missouri Public Service Commission P.O. Box 360 Jefferson City, MO 65102 s
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J. Walker O. Maynard (KG&E)
J. H. Smith (Bechtel)
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ULNRC-1442 RESPONSES TO ENCLOSURE 2 REQUEST FOR ADDITIONAL INFORMATION 1.
The discussions in Sections 2.1 and 2.2 differentiate between identifying SGTR occurrence and identifying which steam generators (SGs) have ruptured tube (s), and suggest indications for these identifications.
However, it is not clear that this distinction between the two different diagnostic activities is considered in the analyses reported. The symptoms you have identified, their alarms, and operator responses may not be appropriate for the event scenario assumptions.
Include explicit consideration of the symptoms, indicating instrumentation, alarms or procedural directives, and operator responses, including times for each in the timetables and analyses for each diagnostic activity. Do this separately for the identification of SGTR occurrence and for the identification of which SG(s) have ruptured.
1.
Response
A.
Identification of SGTR Transient Operator actions in response to a SGTR are assumed to follow plant specific emergency procedures. Depending whether or not a reactor trip had occurred, Off-Normal or Emergency Operating Procedures would direct the operator to identification of a SGTR.
Plant-specific Off-Normal procedures address some actions to take if an SG tube rupture is in progress with no reactor trip or safety injection. With condenser air removal radiation monitors or steam generator blowdown radiation monitors indicating above normal or alarming, the plant chemistry departments would start sampling generators to determine the faulted SG.
Associated instruments, alarms, and setpoints are tabulated in Attachment 1.
l The operator would enter E-0 on a reactor trip or safety injection, I
whether the signal was automatic or a result of manual actuation.
l Through symptom-based diagnosis, the operator is directed to the proper Optimal Recovery Guidelines to facilitate optimal recovery.
As directed in E-0, the operator would review radiation levels in the SG blowdown and/or the condenser air removal systems. Abnormal levels in either of these areas indicates excessive primary-to-secondary leakage and directs the operator to E-3.
Once in E0P E-3, the operator would be directed to identify the ruptured SG(s).
Stuck-Open ARV Scenario For the submitted Stuck-Open ARV case, reactor trip did not occur until nearly 8.4 minutes. Within three minutes of the SGTR, the operator would have observed abnormal radiation levels in the SG blowdown and condenser air removal and would be preparing for reactor trip and SG identification via Off-Normal Procedures (See ).
ULNRC-1442 RESPONSES TO ENCLOSURE 2 REQUEST FOR ADDITIONAL INFORMATION 1.
' Response (Cont'd.)
Overfill Scenario For the submitted Overfill case, reactor trip occurs at 2.4 minutes.
This may not be sufficient tine to identify the SGTR via the Off-Normal Procedures. However, once in E-0 the operator would quickly identify a SGTR via abnormal radiation levels in the SG blowdown and condenser air remaval, or by an increase in SG level-af ter reactor trip (See Attachment 2, step 5).
In either scenario, the operator has several other indications of an 4
SGTR which are not specified in the initial procedures but which may be used to verify his findings. These include:
1.
Uncontrolled SG Level Increase (as seen on control board indications or on level traces recorded prior to trip) i l
2.
Turbine Driven Auxiliary Feedwater Exhaust High Radiation Monitor (identifies either B or C SG) 3.
SG Atmospheric Relief Valve Radiation Monitors 4.
SG Samples i
a.
Conductivity increase due to boron additions to SG water b.
Activity level increase These indications are mentioned since their use has been observed in SGTR simulator exercises.
B.
Identification of the Faulted SG Once the operator has proceeded to E0P E-3, procedural guidance quickly requires identification of the faulted SG.
This is accom-p11shed by observing one of the following (Step 2, E-3):
1.
Unexpected increase in any SG narrow range level, or i
2.
High radiation from any SG sample, or 3.
High radiation from any SG atmospheric relief, or 4.
High radiation from any SG blowdown line.
These indications are obtained by consulting instrumentation within the control room (except for the case of SG manual sampling where the infor-mation is communicated to the control room). High radiation indications are given by both alarms and displays.
ULNRC-1442 RESPONSES TO ENCLOSURE 2 REQUEST FOR ADDITIONAL INFORMATION 1.
Response (Cont'd.)
It should be noted that in both submitted scenarios the loss of off-site power (LOOP) is assumed to be coincident with reactor trip. Since the radiation monitors are not powered by Class IE power supplies, updated information is not available after LOOP. As estimated in the response to question 2, the responsiveness of the required radiation alarms is on the order of 2-3 minutes. Although radiation alarms would occur prior to trip, given adverse impacts from LOOP and the requirement to differentiate between SGs, radiation monitors cannot be assumed for verifying the faulted SG.
Thus identification of the faulted SG must be based on narrow range indication or manual SG liquid sampling.
Overfill Scenario For the overfill case, uncontrolled narrow range rise would be recognized at approximately 6 minutes into the accident. This leaves ample time for operators to complete the isolation steps within 16 minutes (See Attach-ment 2).
Stuck-Open ARV Scenario For the ARV case, level does not recover until 40 minutes after the SGTR.
Therefore, faulted SG identification is achieved " procedurally" via manual SG liquid sampling.
Manual sampling and faulted identification would take 15-20 minutes. Given that sampling is initiated when radiation monitors indicate abnormal levels (<3 minutes), operators have sufficient time to complete isolation activities by 28.4 minutes (See Attachment'2).
Note that emphasis was placed above on " procedural identification".
Operators have available other indications outside the emergency proce-dural guidelines to identify the faulted SG including: initial feed-water / steam mismatch, SG liquid conductivity, SG pressures, and SG wide range indication.
2.
The report references the Westinghouse Emergency Response Guidelines (ERGS) to identify which operator actions are performed in response to SGTR scenarios. There does not seem to be adequate plant specific infor-mation for the ERGS to provide adequate guidance, to adequately identify and quantify SGTR diagnostics. For instance, since the ERGS do not iden-tify radioactivity control as a critical safety function, it is not clear that the operator would be properly directed to consult radiation monitor-ing equipment or that such consultation would be timely.
Identify (1) instrumentation and controls which the analyses assume the operator will use for diagnostic purposes; (2) the procedures that will be signalled by each; (3) alarms or procedural directives which will alert the operator to use them.
Discuss the sensitivity, responsiveness, availability, and qualification of these instruments and controls.
ULNRC-1442
-RESPONSES TO ENCLOSURE 2 REQUEST FOR ADDITIONAL INFORMATION 2.
Response
Although SG radioactivity control is not identified as a Criticality Safety Function, it is treated in Off-Normal and Emergency Operating Procedures as discussed in the previous response.
As outlined in Attachment 1, several indications are available for diagnosing a SGTR. These indications include:
1.
SG Narrow and Wide Range Level, 2.
SG Pressure, 3.
Feed / Steam Flow, 4.
SG Blowdown Radiation Monitors, 5.
Condenser Air Discharge Monitor, 6.
SG Liquid Radiation Monitor, and 7.
Turbine AFW Exhaust Radiation Monitor.
After the SGTR is diagnosed, the following instruments may be used to identify the faulted SG:
1.
SG Narrow and Wide Range Level, 2.
SG Pressure 3.
Feed / Steam Flow, 4.
Manual SG Liquid Sampling by Chemistry Department, and 5.
The use of these instruments in the postulated SGTR scenarios has already been discussed in detail in the response to question 1.
For those instruments used for diagnosis in the submitted analyses, the sensitivity, responsiveness, availability, and qualification of these instruments are addressed in the Table 1.
The procedures signalled by abnormal indications and the procedural steps in which the above controls were used are repeated in Table 1 for completeness.
The bulk of this information was taken from the SNUPPS FSAR; references are included where appropriate.
ULNRC-1442 RESPONSES TO ENCLOSURE 2 REQUEST FOR ADDITIONAL INFORMATION 3.
The discussion of Section 2.0 refers to operator actions based on ERG Section E-3, SGTR; however, this section does not discuss event mile-stones and operator actions prior to entry into the ERGS and actions based on ERG Section E-0.
Address this portion of the SGTR scenario.
3.
Response
Milestones and operator actions prior to entry into the ERGS and actions based on E-0 were addressed in response to question 1.
4.
The operator actions assumed in the analyses of the report are predicated on a predetermined course of operator action which is not detailed or justified. Observations reported in WCAP-10599 indicate that during the ERG validation program operator uncertainty and incorrect interpretations have occurred.
Specific incidents cited in this report are particularly applicable to SGTR events. Justify for the SGTR Analysis scenarios the assumed course of operator actions, given NUREG-0800 assumptions and their consequent indications and ERG-instructed responses to those potential indications.
4.
Response
The event symptoms, alarms, operator actions, and times have been itemized in Attachment 2 for the postulated scenarios.
For each scenario the operator's course of action is justified by procedural guidance as outlined in Attachment 2.
Attachments 3, 4, and 5 are simplified versions of the major procedures the operaters will use during the transient. These attachments are based on Callaway Plant procedures. Wolf Creek Plant procedures are very similar, with only minor variance in step numbering and cautionary notes.
In reference to operator uncertainty and incorrect interpretations reported in WCAP-10599, the majority of errors were attributed to limited procedural training and operator unfamiliarity with the control board. " Wrong-column" errors which are human-factor related were made infrequently (10 out of 2452 possible steps).
Since the WCAP-10593 verification and validation exercises, the Westinghouse ERGS have been developed into Callaway and Wolf Creek procedures. The application of these procedures in operations and training exercises has resulted in decreased incidence of errors and enhanced effectiveness.
Indeed, the Revision 0 verification and validation of the SNUPPS generic procedures on the Callaway simulator showed that the procedures were very effective in limiting human errors.
Given Callaway and Wolf Creek operators' experience and familiarity with their equipment, incorrect actions and their consequences should be minimal.
If errors are made, the redundant checks, independent safety status monitoring, and fundamental similiarity in recovery steps would provide ample opportunity to correct any errant actions. -
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ULNRC-1442 RESPONSES TO ENCLOSURE 2 REQUEST FOR ADDITIONAL INFORMATION 4.
Response (Cont'd.)
For an SGTR transient, several steps in the procedures ensure an operator's return to the proper action steps if an operator has progressed into the wrong recovery procedure.
Emergency procedure ES-0.0, Re-Diagnosis, allows the operator to determine or confirm the most appropriate recovery procedure.
Emergency Procedure E-0, Reactor Trip or Safety Injection, branches to Emergency Procedure E-3, SGTR, four separate times. The Functional Restoration Guidelines (FRGs) are entered as a result Critical Safety Functions monitoring by the operating crew.
In the case of an SGTR, SG high level indications vould lead the crew to FRG-H.3 and then back to E0P-3.
Given these considerations, operator errors are not expected to significantly impact the mitigation of the transient.
Since recognition of an error would occur promptly, operator action times assumed in the analyses are enveloped.
5.
Fcenarios postulated in the SNUPPS SGTR analyses presume the identification by the operator that a SGTR event is in progress and that he has transitioned to the E-3 procedure. This presumption is not adequately justified for a NUREG-0800 scenario.
For a NUREG-0800 scenario itemize step-by-step, from time of tube rupture to time of event termination, i.e.,
cold shutdown, all events accompanying symptoms, alarms, operator actions, and times associated with each. This description should include details prior to entry into the ERGS and all transitions in the ERGS. All operator behavior should be justified, including assumptions that the operator would not make erroneous transitions.
Assumptions required by NUREG-0800, e.g.,
loss of offsite power, stuck rod, and their impact on operator actions should be considered. Also, other activities appropriate to operation during SGTR scenarios (e.g., interaction with Emergency Plan Emergency Action Levels) should be accounted for.
5.
Response
The event symptoms, alarms, operator actions, and times have been itemized on Attachment 2 for the postulated scenarios.
For each scenario the operator actions are justified by procedural guidelines. As noted in response to question 4, the inherent structure of the emergency procedures with redundant checks, independent safety status monitoring, and fundamental similarity in recovery steps provides several redundant steps to correct any incorrect actions.
Assumptions required by NUREG-0800 have been considered in the course of the transient.
In the case of loss of off-site power coincident with reactor trip, updated radiation monitoring is not accounted for after the trip. The assumption that the highest worth control rod is stuck in the fully withdrawn position has no effective impact on the SGTR scenarios, since there is no fuel failure.
ULNRC-1442 RESPONSES TO ENCLOSURE 2 REQUEST FOR ADDITIONAL INFORMATION 5.
Response (Cont'd)
In reference to the Radiation Emergency Response Plan (RERP) actions, implementation of the RERP is the responsibility of tha Emergency Coordinator (Duty Emergency Director at Wolf Creek). Although he has overall responsibility for the plant, he is not directly involved in the recovery.
Implementation of the Emergency Operating procedures will be by the on-shift crew. The event will be classified and the necessary notifications will be made in accordance with the plant specific approved RERP. The Emergency Classification would probably be either an ALERT or a SITE EMERGENCY (SITE AREA EMERGENCY at Wolf Crek).
In any event, the RERP does not effect operator recovery actions.
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ULNRC-1442 TABLE 1 INSTRUMENTS AND RADIATION MONITORS USED FOR SGTR DIAGNOSIS (NOTE 7)
S2n:sitivity Procedural Step MDC (uCi) Range (uCi)
Monitor Requiring Use(8) cc cc Responsiveness Availability Qualification References
-6
-7
-2 Radn Monitor E-0 Steps 23b, 1x10 10 to 10 2-3 minutes 11.5-2 (BM-RE-52) 29b E-3 Step 2d
-6
- 2. SG Blowdown Note 1 1x10 10~
to 10~
Note 3 Notes 4, 5 NSR/NS FSAR Table Proc. System E-0 Steps 23b, 2-3 minutes 11.5-2 Monitor 29b (BM-RE-25)
E-3 Step 2d
~2
- 3. Condensor Note 1 2x10~
10~
to 10 Note 3 Note 4, 5 NSR/NS FSAR Table Air E-0 23a 2-3 minutes 11.5-3 Discharge E-0 29a Monitor (GE-RE-92)
- 4. Manual Note 2 Not Applicable 15-20 minutes Note 6 Not Not Blowdown E-3 Step 2b Applicable Applicable Sampling ao' by Chem Department Procedural Step Requiring Sensitivity Instrument Use (8)
Accuracy (% Range) Range Responsiveness Availability Qualification 5.
- 1. SG Narrow E-0 Step 28a i 4%
0-100%
< 2 seconds 4 channel Class IE-FSAR Table Range E-3 Step 2a available,
- pressure, 7.5.1 Level I required temperature, 7.5.2, and A: AE-L-per SG radiation, 3.11(B) 517,518, spray, etc.
519,551 CB SC, LP (see FSAR B: AE-L-3.11(B))
527,528, 529,552 C: AE-L-537,538, 539,553 D: AE-L-547,548, 549,554 NSR - Non-Safety Related Component; NS - Non-Siesmic Qualification; CB-Control Board (Main);
SC - Systems Cabinets in Control Room; LP - Local Panel
ULNRC-1442 NOTES TO TABLE 1 NOTE 1:
On alarm or alert, the operator would consult plant-specific alarm response procedures. The alarm response procedures direct the operator to Off-Normal Procedures.
NOTE 2:
Various steps in the plant-specific alarm response procedures require notification of the plant chemistry department for manual sampling and/or evaluation.
NOTE 3:
Local microprocessor receives pulse signal from the detector.
It processes the signal and transmits a one minute average to the RRIS computer. The control room computer polls the whole system every 2 seconds through 2 independent chains. Data can be displayed for present, 10 min, I hour, 1 day, and monthly averages.
NOTE 4:
The operator has this information available via control room digital radiation display SP-056A.
NOTE 5:
Power supply is Non-1E. Monitor lost on loss of off-site power.
NOTE 6:
Allows differentiation between individual SG(s).
l NOTE 7:
The parameters listed are defined as follows:
Sensitivity - the accuracy and range over which the instrument operates.
Responsiveness - delays associated with receipt of signal (process or electronic).
Availability - operability and accessibility of the instrument.
Qualificaiton - conditions under which instrument was tested and remained operable.
NOTE 8:
Procedural steps based on Callaway E0Ps. Wolf Creek procedures are very similar, with only minor variance in step numbering and cautionary notes.
l l
ULNRC-1442 RESPONSES TO ENCLOSURE 2 REQUEST FOR ADDITONAL INFORMATION ATTACHMENT 1 INDICATIONS The operator has several indications of an SGTR. Since primary-to-secondary, leakage adds additional inventory to the ruptured SG, the faulted SG 1evel may be higher than normal prior to trip. This change in' level can be seen in the recorded levels and can be utilized after trip by reviewing these records.
This information assists in confirming an SGTR and also is useful in identifying the affected SG(s). After reactor trip, the level in the ruptured SG should return to narrow range significantly earlier than the unaffected SGs..
In addition, the following annunciators would provide the operators with feedback on the progression of the accident:
Annunciator Title Setpoint 108B S/G "A" LEV DEV 5% Program 109B S/G "B" LEV DEV 15% Program 110B S/G "C" LEV DEV 5% Program i
IllB S/G "D" LEV DEV 15% Program 5
j 108C S/G "A" FLOW MISMATCH 7x10 1bm/hr Mismatch 5
109C S/G "B" FLOW MISMATCH 7x10 1bm/hr Mismatch 110C
'S/G "C" FLOW MISMATCH 7x10'1bm/hr Micmatch 5
111C S/G "D" FLOW MISMATCH 7x10 1bm/hr Mismatch 1
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1 ATTACHMENT 1 - -
ULNRC-1442 RESPONSES TO ENCLOSURE 2 REQUEST FOR ADDITIONAL INFORMATION ATTACHMENT 1 INDICATIONS Although these symptoms will be evident soon after reactor trip for a large tube rupture, the SG 1evel response may not be noticeably different from normal for small ruptures or leaks.
In that case, high radiation indications may be necessary for positive identification of the ruptured SG.
Alarms associated with high radiation indications are listed below:
RM-11 Chan Monitor Process Alarm Setpoint 111 AB-RE-111 Steam Line "A" 40 mr/hr 112 AB-RE-112 Steam Line "B" 40 mr/hr 113 AB-RE-113 Steam Line "C" 40 mr/hr 114 AB-RE-114 Steam Line "D" 40 mr/hr
-0 256 BM-RE-25 S/G Blowdown 1x10 uc/ml 526 BM-RE-52 S/G B/D Disch Pumps Variable
~4 506 FB-RE-50 Aux Stm Cond Recovery lx10 uc/ml 381 FC-RE-385 TDAFP Exh 40 mr/hr
-5 925 GE-RE-92 Condenser Air Disch 2x10 uc/mi
~0 026 SJ-RE-02 S/G Letdown 1x10 uc/mi The Safety Assessment System (SAS) can provide information to detect a leak.
This information aid is packaged in displays (Example: SGTR) which present critical parameters for the accident.
The SGTR parameters are:
1.
RCS Pressure 2.
Pressurizer Pressure 3.
Containment Temperatyre 4
Containment Pressure, 5.
Containment Humidity 6.
Containment Sump Level 7.
Cordenser Air Discharge Radiation 8.
SG Blowdown Radiation 9.
Highest SG Level 10.
SG Levels (30 minute graph and numerical form)
SG Pressure The balance of Plant computer can display various points or trends which could serve to identify the problem.
Note Presence of these indications may be indicative of steam line break or LOCA as opposed to an SGTR.
ATTACHMENT 1 -
ULNRC-1442 RESPONSES TO ENCLOSURE 2 REQUEST FOR ADDITIONAL INFORMATION ATTACHMENT 2 Analysis Steps for SGTR - Overfill Case 2
Time After Assumed Operator Action SGTR E0P y
Analysis Step to Support Step (Minutes)
Guidance 1.
SGTR None 0.
2.
High Radiation None 2-3 Indications SG Blow-down/ Condenser Air Removal 3.
Rx Trip on OT T A.
Go to E-0, 2.4 3
4.
LOOP None 2.4 no steam dump RCPs tripped FW terminated 5.
SGTR Identified Either:
A.
Identify NR in one SG 6-10 E-0 Step 28 increasing, uncon-
- trolled, or B.
Identify high radia-8-10 E-0 Step 22 tion indications in or 29 Blowdown and Condenser Air Removal AND C.
Proceed to E-3 E-0 Step 28, 22 or 29 6.
Isolate Faulted SG A.
Identify an unexpected
<16 E-3 Step 2a increase in a SG NR level B.
Check ruptured SG(s)
<16 E-3 Step 3c atmospheric steam dump - CLOSED C.
Terminate AFW 16.
E-3 Step 4b Note 1:
Procedural steps based on Callaway E0Ps. Wolf creek procedures are very similar with only minor variance in step numbering and cautionary notes.
Note 2:
Times given by ranges or limiting values represent actions which required no physical manipulation and/or was not a direct analysis step. Times given by decimal numbers indicate analysis steps in which an action occurred to change the course of the transient.
f ATTACHMENT 2 i
ULNRC-1442 RESPONSES TO ENCLOSURE 2 REQUEST FOR ADDITIONAL INFORMATION ATTACHMENT 2 Analysis Steps for SGTR - Overf111 Case Time After Assumed Operator Action SGTR E0P Analysis Step to Support Step (Minutes)
Guidance 7.
Check intact SG 1evel A.
Control AFW flow to 19.2 E-3 Step 7b maintain NR between 4-50%.
8.
RCS Cooldown A.
Check Ruptured SG
<24 E-3 Step 13 Pres >615 psig B.
Determine required
<24 E-3 Step 14a RCS temp C.
Initiate steam dump 24.
E-3 Step 14b using ARV D.
Verify RCS wide
<31 E-3 Step 14e range temp less than required temperature E.
Stop RCS cooldown 31.
E-3 Step 14d (close ARV) 9.
RCS Depressurization A.
Check ruptured SG
<34 E-3 Step 15 pressure B.
Check RCS Subcooling
<34 E-3 Step 16 (compare core exit temp to wide range pressure)
C.
Depress. using PZR 34.
E-3 Step 18a PORV D.
Close PZR PORV when 35.
E-3 Step 18b RCS Pressure = SG Pressure
- 10. SI Termination A.
Check RCS Subcooling
<38 E-3 Step 20a (compare core exit temp to wide range pressure)
ATTACllMENT 2.. -
ULNRC-1442 RESPONSES TO ENCLOSURE 2 REQUEST FOR ADDITIONAL INFORMATION ATTACHMENT 2 Analysis Steps for SGTR - Overfill Case Time After Assumed Operator Action SGTR E0P Analysis Step to Support Step (Minutes)
Guidance B.
Check NR Level in
<38 E-3 Step 20b Intact >4%
C.
Check RCS Wide Range
<38 E-3 Step 20c Pressure Increasing D.
Check PZR Level >5%
<38 E-3 Step 20d E.
Stop SI Pumps 38.
E-3 Step 21a F.
Stop all but one CCP 38.
E-3 Step 21b
- 11. Pressure Re-Equalita-A.
Equalize pressures 43.
E-3 Step 29 tion
- 12. Post-SGTR Cooldown A.
Go to ES-3.1 or E-3 Step 38 ES-3.2 B.
Initiate RHR Cooling 120.
ES-3.1 or ES-3.2 ATTACifMENT 2 ULNRC-1442 RESPONSES TO ENCLOSURE 2 REQUEST FOR ADDITIONAL INFORMATION ATTACHMENT 2 Analysis Steps for SCTR - Stuck-Open ARV Time After Assumed Operator Action SGTR EOP Analysis Step to Support Step (Minutes)
Guidance 1.
SGTR None 0.
2.
High Radiation A.
Operator coordinates 2-3 Plant Indications in SG with Chemistry Depart-Specific Blowdown / Condenser ment to sample SG for Off-Normal Air Removal activity Procedures 3.
Rx Trip on OTAT A.
Go to E-0.
8.4 4.
LOOP None 8.4 no steam dump RCPs tripped FW terminated 5.
Faulted SG ARV None 8.4 Fails Open 6.
SGTR Identified A.
Identify high radia-8-10 E-0 Step 22 tion alarms in Blow-or 29 down and condenser
~
Air Removal B.
Proceed to E-3 7.
Isolate Stuck-ARV A.
Equipment Operator
<16 E-3 Step 3c dispatched to manu-ally isolate failed ARV 8.
Identify Faulted SG A.
Identify via SG 18-23 E-3 Step 2 sampling results (contact with Chemistry) 9.
Isolate Faulted SG A.
Terminate AFW 16.
E-3 Step 4b B.
Failed ARV isolated 28.4 Note 3: AFW terminated to faulted SG prior to 28.4 minutes in order to maximize off-site dose.
ATTACHNENT 2 ULNRC-1442 RESPONSES TO ENCLOSURE 2 REQUEST FOR ADDITIONAL INFORMATION ATTACHMENT 2 Analysis Steps for SGTR - Stuck-Open ARV.
Time After Assumed Operator Action SGTR E0P Analysis Step to Support Step (Minutes)
Guidance
- 10. RCS Cooldown A.
Check Ruptured SG
<40 E-3 Step 13 Pres >615'psig B.
Determine required
<40 E-3 Step 14a RCS temp C.
Initiate steam dump 40.
E-3 Step 14b using ARV D.
Verify RCS wide
<50 E-3 Step 14e range temp less than required temperature E.
Stop RCS cooldown 50.4 E-3 Step 14d (close ARV)
- 11. RCS Depressurization A.
Check ruptured SG
<53 E-3 Step 15 pressure B.
Check RCS Subcooling
<53 E-3 Step 16 (compare core exit temp to wide range pressure)
C.
Depress, using PZR 53.4 E-3 Step 18a PORV D.
Close PZR PORV when 55.
E-3 Step 18b RCS Pressure = SG Pressure
- 12. SI Termination A.
Check RCS Subcooling
<58 E-3 Step 20a (compare core exit temp to wide range v
pressure)
B.
Check NR Level in
<58 E-3 Step 20b Intact SCs >4%
e ATTACHMENT 2 ULNRC-1442 RESPONSES TO ENCLOSURE 2 REQUEST FOR ADDITIONAL INFORMATION r
ATTACHMENT 2 An'alysis Steps for'SGTR - Stuck-Open ARV Time After Assumed Operator Action SGTR E0P Analysis Step to Support Step (Minutes)
Guidance C.
Check RCS Wide Range
<58 E-3 Step 20c Pressure Increasing D.
Check PZR Level >5%
<58 E-3 Step 20d E.
Stop SI Pumps 58.
E-3 Step 21a F.
Stop all but one CCP 58.
E-3 Step 21b
- 13. Pressure Re-Equaliza-A.
Equalize pressures 63.
E-3 Step 29 tion
- 14. Post-SGTR Cooldown A.
Go to ES-3.1 or E-3 Step 38 ES-3.2
'y B.
Initiate RHR Cooling 120.
ES-3.3 s
k s
o 8
ATTACHMENT 2.-
ULERC-1442 E-0 REACTOR TRIP OR SAFETY INJECTION Rev. 3 EMERGENCY OPERATING PROCEDURE E-0 REACTOR TRIP OR SAFETY INJECTION A.
PURPOSE This procedure provides instructions to verify proper recponse of the automatic protection systems following manual or auto-matic actuation of a reactor trip or safety injection, to assess plant conditions, and to identify the appropriate recovery procedure.
NOTE:
Steps 1 through 14 are IMMEDIATE ACTION steps ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 1.
Verify Reactor Trip 2.
Verify Turbine Trip 3.
Verify Power to NB01/NB02 4.
Check If SI IS Actuated NOTE:
Be aware that ES-0.0, REDIAGNOSIS, is available to help provide direction in determining proper accident analysis.
5.
Ensure Feedwater Isolation l
l 6.
Ensure CISA 7.
Ensure AFW Actuation 8.
Ensure SI Initiation 9.
Ensure One
'CW Pump Running in Each Train 10.
Ensure ESW Pumps -
RUNNING ATTACHMENT 3 ULNRC-1442 E-0 REACTOR TRIP OR SAFETY INJECTION Rev. 3 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 11.
Ensure Containment Coolers - RUNNING IN SLOW SPEED 12.
Ensure CPIS 13.
Check If Main Steam-lines Should Be Isolated 14.
Ensure Containment Spray Not Required i
15.
Ensure Safety Injection Flow 16.
Ensure AFW Flow -
GREATER THAN 260,000 LBM/HR 17.
Ensure AFW Valve Align-ment - PROPER EMERGENCY t
ALIGNMENT 18.
Ensure SI Valve Align-ment - PROPER EMERGENCY i
ALIGNMENT 19.
Check RCS Average Temp-erature - STABLE AT OR TRENDING TO 557 DEG.
F.
20.
Check PZR PORVs And Spray Valves 21.
Check If RCPs Should Bo Stopped 2 2.-
Check If SGs Are Not Faulted J
k i
ATTACHMENT 3 -
ULNRC-1442 E-0 REACTOR TRIP OR SAFETY INJECTION Rev. 3 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 23.
Check If SG Tubes Are Not Ruptured a.
Condenser Air a.
Go To E-3, (STEAM Removal Radiation -
GENERATOR TUBE NORMAL RUPTURE), Step 1.
b.
SG Blowdown Radia-b.
Go to E-3, (STEAM tion - NORMAL GENERATOR TUBE RUPTURE), Step 1.
24.
Check If RCS Is Intact 25.
Check If SI Flow Should Be Reduced c.
RCS Pressure - STABLE c.
DO NOT STOP ECCS OR INCREASING PUMPS.
Go to Step 27.
d.
PZR level - GREATER d.
DO NOT STOP ECCS THAN 5%
PUMPS.
27.
Implement CSF-1 (CRITICAL
-SAFETY FUNCTION STATUS TREES)
CAUTION:
Alternate water sources for AFW pumps will be necessary if CST level decreases to less than 15%.
28.
Check SG Levels b.
IF narrow range Tevel in any SG b.
Control feed flow to continues to maintain narrow range increase in an level between 4% and 50%.
uncontrolled manner, THEN go to E-3, (STEAM GENER-ATOR TUBE RUPTURE), Step 1.
29.
Check Secondary 29.
Go to E-3 (STEAM GENER-Radiation - NORMAL ATOR TUBE RUPTURE),
Step 1.
ATTACHMENT 3 ULNRC-1442 ES-0.0
'REDIAGNOSIS Rev. 1 EMERGENCY OPERATING PROCEDURE ES-0.0 REDIAGNOSIS
' A.
PURPOSE This procedure provides a mechanism to allow the operator to determine or confirm the most appropriate post accident recovery procedure.
ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 4
1.
Check If Any SG Is Intact 2.
Check If All SGs Are Intact 3.
Check If SG Tubes Are Ruptured a.
ANY SG LEVEL INCREASING IN AN UNCONTROLLED MANNER
-OR-b.
ANY SG WITH HIGH RADIATION 4.
You Should Be In An E-3 i
Or ECA-3 Series Procedure i
l i.
l l
l l
l l
1 ATTACHMENT 4..-
ULNRC-1442 E-3 STEAM GENERATOR TUBE RUPTURE Rev. 3 EMERGENCY OPERATING PROCEDURE E-3 STEAM GENERATOR TUBE RUPTURE A.
PURPOSE This procedure provides actions to terminate leakage of reactor coolant into the secondary system following a steam generator tube rupture.
ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 1.
Check If RCPs Should Be Stopped 2.
Identify Ruptured SG(s) 2.
Continue with Steps 5 through 12.
WHEN a.
Unexpected increase ruptured SG(s) in any SG narrow identified.
THEN range level perform Steps 3 & 4.
-OR-b.
High radiation from any SG sample
-OR-c.
High radiation from any SG Steamline
-OR-d.
High radiation from any SG blowdown line 3.
Isolate Flow From Ruptured SG(s) c.
Check ruptured SG(s) atmos-c.
IF atmospheric pheric steam dump - CLOSED steam dump can NOT be closed THEN locally isolate atmospheric steam dump.
4.
Check Ruptured SG(s) Level b.
Control feed flow to b.
IF narrow range maintain narrow range level still level between 4%.
increases, THEN throttle feed flow to zero or isolate feed flow to ruptured SG(s).
ATTACHMENT 5 ULNRC-1442 E-3 STEAM GENERATOR TUBE RUPTURE Rev.-3 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 5.
Check PZR PORVs And Block Valves 6.
Check If SGs Are-Not Faulted 7.
Check Intact SG Levels b.
Control feed flow to maintain narrow range level between 4% and 50%.
8.
Reset SI 9.
Reset CISA and CISB 10.
Establish Instrument Air
)
To Containment 11.
Verify All AC Buses -
ENERGIZED BY OFFSITE POWER 12.
Check if RHR Pumps Should Be Stopped 13.
Check Ruptured SG(s)
Pressure - GREATER THAN 615 PSIG 14.
Initiate RCS Cooldown l
a.
Determine required RCS temperature.
I b.
Dump steam to condenser b.
Manually or from intact SG(s) at locally dump steam maximum rate.
at maximum rate from intact SG(s)
(1)
Use atmospheric steam dumps L
-OR-l (2)
Dump steam through TD-AFP l
on RECIRC.
l c.
Core exit TCs or RCs wide
[
range RTD's - LESS THAN l
REQUIRED TEMPERATURE j
d.
Stop RCS cooldown, l-i I
ATTACHMENT 5,
m -
,--,-,-,.,-,-.v
,m..,
r%,,n-----------r-
.em~
---,-,-e-----T
=n-
--o----
ULNRC-1442
> n E-3 STEAM GENERATOR TUBE RUPTURE Rev. 3 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 15.
Check Ruptured SG(s)
Pressure - STABLE OR INCREASING 16.
Check RCS Subcooling -
PRESSURE AND TEMPERATURE WITHIN PERMISSABLE RANGE i
17.
Depressurize RCS To Minimize Break Flow and Refill PZR 18.
Depressurize RCS Using PZR PORV To Minimize Break Flow And Refill PZR
!~
a.
PZR PORV - AT LEAST ONE AVAILABLE
[
b.
Open one PZR PORV until ANY of the following conditions satisfied:
7 (1)
BOTH of the following (a) RCS pressure -
LESS THAN l.
RUPTURED SG (s) i PRESSURE i
-AND-(b) PZR level - GREATER THAN 5%
-OR-(2)
PZR level - GREATER THAN 80%
(3)
RCS subcooling f
c.
Close PZR PORV l
19.
Check RCS Pressure -
INCREASING i
ATTACHMENT 5....
ULNRC-1442 E-3 STEAM GENERATOR TUBE RUPTURE Rev. 3 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 20.
Check If SI Flow Should Be Terminated a..-RCS subcooling b.
Secondary heat sink 4
c.
RCS Pressure - STABLE OR INCREASING 21.
Stop ECCS Pumps And Place In Standby a.
SI Pumps b.
All but one CCP 22.
Establish 70 GPM Charging Flow 23.
Isolate BIT 24.-
Control Charging Flow To Maintain Pressurizer Level 25.
Verify ECCS Flow Not Required s
26.
Check VCT Makeup control System 27.
Check If Letdown Can Be Established 28.
Control RCS Pressure And Makeup Flow To Minimize RCS - To - Secondary Leakage r
30.
Check If Diesel Generators Should Be Stopped 31.
Minimize Secondary System Contamination ATTACHMENT 5 _..
-. ~.
ULNRC-1442 E-3 STEAM GENERATOR TUBE RUPTURE Rev. 3 32.
Turn On PZR Heaters As Necessary To Saturate PZR Water At Ruptured SG(s)
Pressure 33.
Check RCP Cooling - NORMAL
'34.
Check If RCP Seal Return Flow Should Be Established 35.
Check RCP Status 36.
Check If Source Range Detectors Should Be Energized 37.
Shutdown Unnecessary Plant Equipment 38.
Go To Appropriate Post-SGTR Cooldown Method a.
Go to ES-3.1, (POST-SGTR COOLDOWN USING BACKFILL), Step 1
-OR-b.
Go to "S-3.2, (POST-SGTR COOLDOWN USING BLOWDOWN), Step 1
-OR-c.
Go to ES-3.3, (POST-SGTR COOLDOWN USING STEAM DUMP), Step 1 ATTACHMENT 5