ML20216F437
| ML20216F437 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 03/10/1998 |
| From: | Hughey W ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GNRO-98-00030, GNRO-98-30, IEIN-97-063, IEIN-97-63, NUDOCS 9803180390 | |
| Download: ML20216F437 (11) | |
Text
e O
Ent y r;tions, Inc.
Pcut Geson. MS 39150 Tel 601437-6470 W.K.Hughey March 10, 1998 gg.m A%ws U.S. Nuclear Regulatory Commission Mail Station P1-37 Washington, D.C. 20555 Attention:
Docurrent Control Desk
Subject:
Grand Gulf Nuclear Station Unit 1 Docket No. 50-416 License No. NPF,29 Revised Request for Alternative to 50.55a Examination Requirements of Category B1.11 Reactor Vessel Welde. Relief Request 1-2-00001 GNRO-98/00030 Gentlemen:
In 1995 the Boiling Water Reactor Vessel & Intemals Project (BWRVIP) transmitted BWR /IP-05, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations" to the Nuclear Regulatory Commission (NRC). The NRC's evaluation prompted numerous industry /NRC meetings and the issuance of NRC Information Notice 97-63.
NRC issued IN 97-63 to inform addressees of the Status of NRC staff's review of "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)." Because of additional analysis requested by the NRC, the staff indicated in the information notice that consideration will be given to technically-justified requests for relief from the augmented examination in accordance with 10 CFR 50.55a(a)(3)(i),10 CFR 50.55a(a)(3)(ii), and 10 CFR 50.55a(g)(6)(ii)A(5) for BWR licensees who are scheduled to perform inspections of the BWR RPV circumferential shell welds during the Fall 1997 or Spring 1998 outage seasons. Further, acceptably-justified reliefs would be considered for inspection delays of up to two operating cycles for BWR RPV circumferential shell welds only.
Grand Gulf Nuclear Station (GGNS) hereby requests relief from the aforementioned requirements. The attached relief request is a revision to and supersedes GNRO-98/00015 dated February 11,1998. The relief request has been prepared in accordance with the guidance of IN 97-63 using the BWP. VIP suggested format. After telephone discussions with the NRC staff, the original relief request has been modified to provide a GGNS specific g'
system analysis for the probability of a cold overpressure transient. The conclusions from I
f our original request, as provided in our previous submittal, have not been modified by this revision. This attemative is being requested under the provisions of 10 CFR 50.55a(a)(3)(i) and will provide an acceptable level of quality and. stety. The Inservice Inspection 9003180390 980310 l
- * ' ' ' !hlhh h PDR ADOCK 05000416 I
G PDR s
r GNRO-98/00030 Page 2 of 3 l
Program at GGNS complies with the 1992 edition with portions of the 1993 addenda of the ASME Boiler and Pressure Vessel Code,Section XI.
GGNS requests your review and approval by April 3,1998 to support GGNS's refueling.
J outage scheduled to commence April 11,1998. Thank you in aavance for your prompt attention in this matter. Should you have any questions or need any additional information, please contact Bill Brice at 601-437-6556.
Yours truly,
)
WKH/WBB/MHB attachment:
Relief Request 1-2-00001 i
GNRO48/00030 Page 3 of 3 cc:
Ms. J. L. Dixon-Herrity, GGNS Senior Resident (w/a)
Mr. L. J. Smith (Wise Carter) (w/a)
Mr. N. S. Reynolds (w/a)
Mr. H. L. Thomas (w/o)
Mr. E. W. Merschoff (w/a)
Regional Administrator U.S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza DrNe, Suite 400 Arlington, TX 760t1 Mr. J. N. Donohew, Project Manager (w/*)
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 13H3 Washington, D.C. 20555 1
l l
I l
ENTERGY OPERATIONS,INC.
GRAND GULF NUCLEAR STATION i
2 nd TEN YEAR INTERVAL REQUEST NO. I-2-00001 1.
COMPONENT / EXAMINATION IDENTIFICATION:
Code Class:
1
References:
ASME Section XI,1992 Edition,IWB-2500; 10 CFR 50.55a(a)(3)(i) and 10 CFR 50.55a(g)(6)(ii)(A)(2);
BWRVIP-05, "BWR Reactor Pressure Vessel Shell Weld Inspection";
NRC Information Notice 97-63, " Status of NRC Staff's Review of BWRVIP-05";
BWRVIP Response to NRC RAI on BWRVIP-05, 12/22/97 Examination Category:
B-A Ite m N o.:
Bl.11 Examination Required:
Volumetric Examination of Welds and Adjacent Base Materials
==
Description:==
Circumferential Shell Welds in Reactor Vessel Component Number:
Q1B13D001 11.
REQUIREMENTS:
ASME Section XI,1992 Edition, IWB-2500 requires the subject welds and associated base material to be volumetrically examined once each interval. The examinations are to be dispersed over the three periods of the interval within the limits specified by IWB-2412-1. Deferral of the examinations until the end of the interval is permissible; however, the examinations during the initial interval were not deferred, and IWB-2420 requires the sequence of examinations established in the first interval to be repeated during subsequent intervals to the extent practical.
j In 1992, Title 10 of the Code of Federal Regulations (10 CFR) were amended with the addition of 50.55a(g)(6)(ii)(A)," Augmented Examination of Reactor Vessel." Section 50.55a(g)(6)(ii)(A)(2) requires licensees to augment their. reactor vessel examinations by implementing once, as part of the insc vice inspection interval in effect on September 8, 1992, the examination requirements for reactor vessel shell welds specified in Item Bl.10 l
of Examination Category B-A," Pressure Retaining Welds in Reactor Vessel"in Table l
IWB-2500-1 of subsection IWB of the 1989 Edition of ASME Section XI, subject to the l
conditions specified in 50.55a(g)(6)(ii)(A)(3) and (4). The augmented examination when I
not deferred in accordance with the provisions of 50.55a(g)(6)(ii)(A)(3), shall be performed in accordance with the related procedures specified in the Section XI Edition Page 1 of 8
and Addenda applicable to the inservice inspection interval in effect on September 8, 1992. For the purpose of this augmented examination, " essentially 100%", as used in Table IWB-2500-1, means more than 90% of the examination volume of each weld, where the reduction in coverage is due to interference by another component or part
- geometry.
Section 50.55a(gX6)(iiXAX3) permits licensees with fewer than 40 months remaining in the inservice inspection interval in etreet on September 8,1992, to defer the augmented reactor vessel examination specified in 50.55a(g)(6)(ii)(A)(2) to the first period of the next inspection interval under certain conditions. However, if the augmented examinations are deferred to the first period of the next inspection interval, 50 55a(g)(6Xii)(AX3)(vi) requires the deferred examinations to be performed in accordance with the related procedures specified in the Section XI edition and addenda l
applicable to the inspection interval in which the augmented examination is performed.
Section 50.55a(g)(6)(ii)(A)(4) indicates that the requirement for augmented examination 4
of the reactor vessel may be satisfied by an examination of essentially 100% of the reactor shell welds specified in 50.55a(g)(6)(ii)(A)(2) that have been completed, or are scheduled for implementation with a written commitment, or are required by 50.55a(g)(4)(i), during the inservice inspection interval in effeo on September 8,1992.
III.
BASIS FOR ALTERNATIVE Pursuant to the provisions of 10 CFR 50.55a(a)(3)(i), and consistent with information contained in NRC Information Notice 97-63, an alternative is requested from the examination of RPV circumferential welds as required by ASME Section XI, IWB-2500, Examination Category B-A, Item No. Bl.11, and 10 CFR 50.55a(g)(6)(ii)(A)(2) as described within. This proposed alternative is to postpone the examination of the PJV circumferential welds for two operating cycles, until Refueling Outage RFl 1 that is presently scheduled to begin in approximately April,2001. The basis for this request for alternative is documented in the report "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)" that was transmitted to the NRC in September 1995 and BWRVIP Response to NRC RAI on BWRVIP-05 that was transmitted to the NRC on December 18,1997.
The BWRVIP-05 report provides the technical basis for eliminating examinations of BWR RPV circumferential shell welds. The BWRVIP-05 report concludes that the probability of failure of the BWR RPV circumferential shell welds is orders of magnitude lower than that of the axial shell welds. Additionally, the NRC assessment demonstrated that examination of BWR RPV circumferential shell welds does not measurably affect the probability of failure. Therefore the NRC evaluation appears to support the conclusions ofBWRVIP-05.
The independent NRC assessment utilized the FAVOR code to perform a probabilistic fracture mechanics (PFM) analysis to estimate RPV failure probabilities. Three key assumptions in the PFM analysis are:
j Page 2 of 8 i
I (1) the neutron fluence was that estimated to be end-of-license (EOL) mean fluence, I
(2) the chemistry values are mean values based on vessel types, and (3) the potential for beyond-design-basis events is considered.
i Although BWRVIP-05 provides the technical basis supporting the request for alternative, the following information is provided to show the conservatisms of the NRC analysis for the Grand Gulf Nuclear Station (GGNS). For plants with RPVs fabricated by Chicago Bridge & Iron (CB&I), the mean EOL neutron fluence used in the NRC PFM analysis 2
was 0.19E+19 n/cm However, the highest surface fluence for the GGNf, RPV beltline 2
region at the end of the requested ahernative period is pred:eted to be 0.102E+19 n/ cm,
Thus the efTect of fluence on embrittlement is much lower, and the NRC analysis as described in the NRC independent assessment is conservative for GGNS in this regard.
Therefore, there is significant conservatism in the already low circumferential-weld-failure probabilities as related to GGNS. Other GGNS RPV shell weld information that j
the NRC staff has requested be included in this relief request is provided in the attached Table 1.
As shown in UFSAR Figure 5.3-9, GGNS does not have any circumferential welds in the beltline region. Ilowever, an evaluation showing the effects of radiation have been performed on the two circumferential welds that are closest to the core. The effects of j
irradiation depicted in this relief request are significantly exaggerated because:
the two welds are not located in the peak fluence region of the beltline, however peak beltline fluence values have been used in their evaluation (weld AB is approximately 5 inches below the core and weld AC is approximately 22 inches above the core),
the fluence used in this relief request represents surface fluence and not 1/4t e
i fluence, and i
there is no credit taken for the attenuation caused by the RPV inner surface e
cladding.
l The results of the evaluations are listed in Table 1. As shown in Table 1, the calculated j
embrittlement shift in RT m (i.e., ART nT) for the GGNS Unit 1 vessel is a maximum of N
N l
22.76 F at the end of the requested relief period. By comparison, using the mean values j
for fluence and weld chemistry assumed for CB&l reactor vessels in Table 7-5 of Enclosure I to the NRC independent assessment report, a ART or f 30.16 F would be N o derived. Therefore, the calculated ARTun7 value for the GGNS vessel is less than, and thus bounded by, the embrittlement shift assumed in the NRC's independent assessment.
Furthermore, it can be seen in the attached Table 1 that the calculated Upper Bound RTsur value for the GGNS near-beltline welds is a maximum of 25.52 F at the end of the requested relief period. For comparison, the Upper Bound RT nT value in Table 7-8 of N to the NRC's independent assessment report of BWRVIP-05 is 32.7 F for fluence reference case 1. Again, the calculated Upper Bound RTun7 values for the GGNS Page 3 of 8
4 vessel circumferential we ~
9ded by the limiting RTmyr from Table 7-8 (CB&l vessels) of the NK
- ssment report, thus providing additional
.,o assurance that the GGNS vosel welds -
'so bounded by BWRVIP-05 report.
An added safety margin has bc:n provided at GGNS by the nondestructive examination (NDE) of the vessel welds. A complete Preservice Inspection (PSI) was performed on all of the RPV shcIl welds, both longitudinal and circumferential, to the maximum extent practical before GGNS initially loaded fuel. The same welds have also completed Inservice Inspection (ISI) ultrasonic examinations required during the first 10-year interval. The examination coverage for both PSI and ISI for all welds except for circumferential weld AA exceeded 90% coverage of the full volume. Weld.AA has been examined over its complete length, but due to scanning limitations from the lower head side of the weld, it was only examined for approximately 67% of the Code required volume.
At the August 8,1997 meeting and in the NRC's independent assessment, the NRC staff indicated that the potential for, and consequences of, nondesign-basis events not addressed in thd BWRVIP-05 report should be considered. In particular, the NRC staff stated that nondesign-basis cold over-pressure transients should be considered. It is highly unlikely that a BWR would experience a cold over-pressure transient. In fact, for a BWR to experience such an event would general:y require several operator errors. At the August 8,1997 meeting, the NRC staff described several types of events that could be precursors to BWR RPV cold over-pressure transients. These were identified as precursors because no cold over-pressure event has occurred at a U.S. BWR. Also at the August 8 meeting, the NRC staffidentified one actual cold over-pressure event that occurred juring shutdown at a non-U. S. BWR. This event apparently included several operator errors that resulted in a maximum RPV pressure of 1150 psi with a temperature range of 79 F to 88 F. As a result of the NRC's concerns, the BWRVIP has included in to their response to the NRC's RAI on BWRVIP-05 significant discussion regarding BWR cold pre',surization events. GGNS has reviewed the BWRVIP's response and concurs that the coaditions and events are accurately depicted and that the procedures and personnel training at GGNS are comparable to those described by the BWRVIP and are adequate to prevent a cold over-pressure transient event. Consequently, the probability of a cold over-pressure transient is considered to be less than or equal to that used in the NRC analysis described in the NRC independent assessment and is conservative for GGNS.
1 Review of Potential Injection Sources That Could Cause a Reactor Pressure Vessel Cold Over-Pressurization:
(
l The Reactor Core Isolation Cooling (RCIC) system is one of the high pressure make-up systems at GGNS. The RCIC system is a steam turbine driven system. RCIC injection during cold shutdown is not possible as no steam is available to drive the RCIC turbine.
The RCIC turbine was designed to also operate on Auxiliary Steam for testing purposes.
The supply line has a removable spool piece and is blind flanged. Operation with Auxiliary Steam is not allowed by procedure.
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t
The Iligh Pressure Core Spray (IIPCS) system is another high pressure make-up system at GGNS. The IIPCS pump is motor operated so it can be operated when the reactor is in cold shutdown. Ilowever, to start the IIPCS system would require either manual initiation, inadvertent initiation or manud wrtup for the llPCS system to start and inject into the reactor vessel. Also, there is a high level interlock for the IIPCS injection valve to prevent overfilling the reactor vessel. This high level interlock cannot be overridden.
Even if the HPCS system is inadvertently started it should not overfill and pressurize the reactor due to the high level interlock.
The Standby Liquid Control (SBLC) is another high pressure system used to shut down the reactor if the control rods fail to insert. The SBLC system has no auto start function so it is unlikely that a spurious initiation could occur. The SBLC system must be manually initiated by a key lock switch. The Plant Supervisor maintains custody of the keys. SBLC is a low flow rate system (about 42 gpm per pump) and is limited to the amount of water that is contained in the storage tank (about 5000 gallons). Even if the SBLC system was manually initiated and not monitored there would not be enough water in the storage tank to fill the reactor from normal water level and would not, therefore, pressurize the reactor.
The Reactor Feed pumps are the high pressure makeup system during normal operation.
The Reactor Feed pumps are steam driven and cannot be operated when the reactor is in cold shutdown because no steam is available to drive the turbine. The Reactor Feed pumps also have a reactor hi level trip.
The Condensate system is the supply source to the Reactor Feed pumps. The Condensate pumps have a discharge pressure of about 150 psig and the Condensate Booster pumps have a discharge pressure of about 650 psig. During operation of both Condensate and Condensate Booster pumps, sufficient temperature margin is provided to ensure that the Technical Specification for the reactor pressure-temperature is not exceeded. This is accomplished by plant procedures dictating when Condensate and Condensate Booster operation is allowed. When the plant is in cold shutdown, reactor temperature is maintained above 70*F per Technical Specifications. If a Condensate pump was started (requires manual action) and lined up for injection and the resulting reactor level increase ignored, the reactor pressure-temperature limit would still not be exceeded since the shut off head of the Condensate pump is about 150 psig (about 300 psig required to exceed limit).
For the reactor pressure-temperature limit to be exceeded, a Condensate pump would have to be manually started, a Condensate Booster pump would have to be manually started, and both manually lined up for injection. Then the injection would have to be ignorci by the operating crew and allowed to continue until the reactor is then pressurized above the pressure-temperature limits. The operating crew would have numerous indications that Condensate was injecting (feedflow indicators and recorders, check valve indication) and reactor level and pressure increases (Upset and shutdown level indication and recorders, narrow and wide range pressure indicators and recorders).
Page 5 of 8
1 Because of the number of operator errors that would have to occur and the number of indications that would have to be ignored, the probability of this event is very low.
The Low Pressure Core Spray (LPCS) system is a low pressure ECCS spray system.
Technical Specifications for the reactor pressure temperature limit permit pressures up to about 300 psig at temperatures from 70 up to 100 F. Above 100 F, pressures permitted
' by Technical Specifications increase immediately to above 700 psig and thereafter increase rapidly with temperature increases. The LPCS system has a discharge pressure of about 500 psig. During refueling outages there is typically only a very short period of time during detensioning and following vessel head retensioning that an overpressurization event could occur. As soon as the vessel head is retensioned, the IOls instruct the operators to begin heatup. Plant procedures also specify that temperatures be maintained between 120 and 130*F during shutdown.- Therefore, the reactor bulk coolant temperature is normally well above 100 F. Procedural controls and the short period of time when the vessel coolant temperatures could be below 100 F make the probability for an over-pressurization due to an inadvertent actuation of this system very low.
The Low Pressure Core Injection (LPCI) systems (3 total) e low pressure ECCS injection systems. The LPCI systems have a discharge pressure of about 300 psig. If they were to be inadvertently initiated or manually sta:ted and lined up to the reactor they would only pressurize the reactor to approximately 300 psig. Technical Specifications requires that reactor metal temperature be maintained above 70 F. Because of this, the Technical Specification requirement for reactor pressure-temperature limit would not be exceeded.
The Control Rod Drive (CRD) system is a high pressure system used to operate control rods. The CRD system is a low flow rate system with about 60 gpm flow rate to the reactor. During cold shutdown conditions reactor water level is maintained with CRD (makeup) and Reactor Water Cleanup (reject). Per plant procedures the reactor head vents are open when reactor coolant temperature is less than 190 F. During cold shutdown conditions the operators closely monitor reactor water level, temperature and pressure.
With the CRD flowrate low and the reactor head vents open, the operators should have sufficient time to react to regain control of reactor pressure, should any abnormalities occur.
Post Outage Primary System Hydrostatic Testing is another postulated over-pressurization event. GGNS has plant procedures as well as Technical Specifications that dictate parameters and steps in performing hydrostatic testing. Hydrostatic testing is considered an " Infrequently Performed Test or Evolution". This requires management oversight, crew briefs, review ofindustry events and assigned responsibilities for the test to be performed. Reactor coolant is heated up to 155 -175 F before reactor pressure is increased to test pressure. Reactor level is maintained with CRD (make-up) and/or RWCU (blowdown) Reactor pressure changes are limited to 50 psi per minute by plant procedures. Two safety relief valves are required to be operable during the test by plant procedures. Because of these strict controls, the likelihood of an overpressurization event during a hydrostatic test is minimal.
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Procedural Controls and Operator Training That Prevent Reactor Pressure Vessel Cold I
Over-Pressurization:
Plant procedures and Technical Specifications dictate bands at which reactor water level, pressure and temperature are to be maintained which ensures an adequate level of safety during all modes of operation. Operation of GGNS follows the steam saturation curve.
Therefore, the operating temperatures are expected to be well in excess of the minimum temperatures required by Technical Specifications. The Control Room Operators are required by procedure to maintain reactor parameters (i.e., water level, pressure and temperature) within these bands and to frequently monitor those parameters. They are also required by procedure to report to the Plant Supervisor anytime operation is outside of a prescribed band. The Plant Supervisor is responsible to ensure that actions are taken to establish those parameters back within the desired band. Also, as previously noted, plant procedures require pre-job briefings and contingency plans before infrequent tests or evolutions are perfomied. Training reinforces these requirements in both classroom and simulator training. Finally, plant conditions, status of plant equipment, special activities along with their potential effect on key plant parameters, and contingency planning are discussed with oncoming crews during shift tumover.
At GGNS, work performed during an outage is scheduled by the Outage Management group. Outage Management includes Senior Reactor Operators who provide oversight of the outage schedule development to avoid conditions that could adversely affect reactor j
water level, pressure or temperature. From the outage schedule, a plan of the day is developed listing the work activities that will be performed that day. The plan of the day schedule is approved and reviewed by management. The plan of the day is assessed for i
shutdown risk to ensure an adequate level of safety is maintained. Any changes to the plan of the day must be approved by management.
The Refueling Integrated Operating Instruction (101) procedure requires that the reactor be depressurized before flooding up to the cold shutdown water level of about 230 inches when entering Refuel operations. Shutdown 101 requires that the Reactor llead Vents be opened when reactor coolant temperature is about 190 F during reactor cooldown. During Ilydrostatic testing, the Reactor Vessel In-Service Leak Test 101 requires reactor coolant temperature be heated up 1o155-175 F and at least 2 Safety Relief Valves to be operable prior to increasing reactor pressure. All of these help ensure the Technical Specifications requirements for reactor pressure-temperature limits are not exceeded.
IV.
CONCLUSION Based on BWRVIP-05, the risk-informed independent assessment performed by the NRC staff, the BWRVIP's response to the NRC's RAI, and the discussion contained within, an alternative to the cited requirements under the provisions of 10 CFR 50.55a(a)(3)(i) to delay the described examinations until Refueling Outage RFl1 is reasonable and will provide an acceptable level of quality and safety.
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4 Table 1 GGNS RPV Shell Weld Information eVariable,
Wariable Value by Weld Sesm Identification !
3, Weld Seam'/i,
AB(LowerCircq f AC(Upper Circ.c
~ AC.(Upper Cire.9 Seem)'.
LSeam)f,
iSeam)% M
> ':.Weiding Process ~
JSAW. !
. SAW h
-SMAWJ' k'
Fluence' @ 13.1 EFPY 0.102 x 10 n/cm2 0.102 x 10 n/cm2 0.102 x 10 n/cm2 (end of relief request period)
Initial RTun7
-40.0 F
- 20.0
- F
- 60.0 F
' Weld Chemistry Factor 41 54 27 Weld Copper Content 0.03 wt%
0.04 wt%
0.02 wt%
0.95 wt%
0.91 wt%
increase in Reference
]
Temperature due to 17.28
- F 22.76
- F 11.38
- F J
Irradiation (ARTun1)
Margin Term 17.28 F 22.76 " F 11.38
- F I
Mean Adjusted Reference Temperature.
- 22.72
- F 2.76
- F
- 48.62
- F (Mean ART)
Upper Bound Adjusted Reference Temperature
- 5.44
- F 25.52
- F
-37.24
- F (Upper Bound ART)
NOTES:-
- 1) GGNS RPV beltline does not contain circumferential welds. Figure 5.3-9 of the GGNS UFSAR shows that weld seam AB is approximately 5 inches below the core and weld seam AC is approximately 22 inches above the core.
- 2) The value is the peak fluence in the beltline region that was linearly interpolated to 13.1 EFPY. The use of peak fluence from the beltline region to compute the shift due to irradiation of the circumferential welds (that are not actually in the beltline region) provides a conservative i
upper bound ARTuor.
l l
I I
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