ML20137G422
| ML20137G422 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 03/12/1997 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20137G377 | List: |
| References | |
| PROC-970312-01, PROC-970312-1, NUDOCS 9704010393 | |
| Download: ML20137G422 (7) | |
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ITYPP-2 INSERVICE INSPECTION PLAN FOR TIIE SECOND TEN-YEAR INTERVAL AT BEAVER VALLEY POWER STATION UNIT #1 RELIEF REQUEST BVI-B3.120-2, REV. O
SUBJECT:
Pressurizer Surge Nozzle Inner Radius Section (RC-TK-t-RADIUS-6)
Prepared by: 5" Md Date: O 2 3 - f0 Supenisor, inspection and Standvds Reviewed by: h D d /~ #AW Date: 3-//-9 7 Director, Nuclear Analysis and Inspon Reviewed by: _3V -0% - \ 9 T Date: #-8 f7 OSC Meeting No.
Reviewed by: O (b'e n k'. ST-1 Date: 2 26-f T .
ORC Meeting No. 1 Approval by: A t_ / .- Date: 3-/2 d Div. VP, Nucleaf Operations / Plant hia' nager 9704010393 97032S PDR ADOCK 05000334 G PDR
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DUQUESNE LIGHT COMPANY Beaver Valley Power Station Unit No.1 RELIEF REOUEST NO. BVI-B3 120-2. Rev. 0 COMPONENT Pressurizer Surge Nozzle Inner Radius Section (RC-TK-1-RADIUS-6)
DRAWING NO.
ISI-E-IP ;
L ASME CODE SECTION XI REOUIREMENT (83S83) 4 Item No. B3.120 (IWB-2500-1, Category B-D) requires volumetric examination.
BASIS OF RELIEF In accordance with 10CFR50.55a(a)(3)(ii), reliefis requested on the basis that compliance with the Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The Beaver Valley Unit 1 pressurizer lower head and the pressurizer surge nozzle is a carbon steel casting to SA-216 GR. WCC. To perform the required UT examination on the surge nozzle inner radius section, the outside surface of the lowei head must be accessible. This surface is made accessible by removing the insulation surrounding the surge nozzle. The design of this insulation requires disconnection of the 78 heater cables from the immersion heaters prior to removing the insulation (see figure 1). Each cable consists of two wires, each mechanically connected to the heater (see figure 2). Great care must be taken during the disconnection to ensure the ceramic terminal blocks to which the heater pins are brazed, are not damaged. Otherwise, an unbrazing/ brazing evolution would be required to replace the blocks. The dose estimate for this exam assumes that no ceramic terminal blocks would require replacement. Another concern involved in this examination is the presence of asbestos in the cablejackets. Though, radioactive contamination is not typically a concern in this area, respirators would be required due to the potential asbestos exposure. Additional cover-alls would be required over the anti-contamination clothing, causing a heat stress concern. Special monitoring and material control would also be necessary due to the presence ofasbestos.
The dose estimate included below is based on a survey obtained on 4/19/96 during 1R11. The hours estimated to perform the activities involved in disconnecting the cables are based on similar efforts performed during 1R08 when two cables were disconnected due to potential short circuits.
Comprehensive dry-run exercises were performed on a mock-up in preparation for the IR08 cfforts.
Therefore, the estimated times used in the dose estimate are believed to be quite accurate.
Page 2 of 4 Relief Reauest No. BVI-B3.120-2. Rev. O If the insulation was removed from this area, the complete code required exam could be performed.
Special search units were designed to perform this specific exammation. The other five pressurizer inner radius sections have been successfully examined. No recordable indications were noted on these examinations. Also, the adjacent safe-end to nozzle weld has been completely examined (IR08) with satisfactory resuhs.
An informal survey of Westinghouse plants found a mix of plants having approved relief requests and others that perform this examination. Those performing this examination have found no unacceptable indications with one exception. One utility found very small cladding cracks. These cracks were attributed to a one time event that was caused by thermal shock when cold water was allowed to enter the empty pressurizer with the heaters on. BV Units 1 and 2 have 30 years of combined operation experience, with no problems in this area. The BV-2 surge nozzle inner radius section and the surge nozzle to vessel weld were UT examined during the current interval and found no recordable indications. The BV-1 surge nozzle inner radius section was not examined in the first interval. Relief was granted based on the configuration of the nozzle and the lack of an adequate UT technique.
7 The radiation exposure for cable disconnection /reconnection, insulation removal / reinstallation, surface preparation, and the UT examination is 54,600 mR as noted in the chart below. The dose estimate is based on a survey conducted with the insulation installed. Once the insulation is removed the rates shown in the survey would increase. Shielding at this location is not practical since the source of the radiation is the component surface to be examined.
DOSE ESTIMATE FOR THE PRESSURIZER SURGE NOZZLE INNER RADIUS SECTION Time (hrs) Individuals Repetitions Dose rate Estimate (mR/hr) (mR)
ELECTRICIANS:
Disconnect heater cables 0.25 1 78 600 11,700 Re-connect heater cables 0.75 1 78 600 35,100 INSULATORS:
Remove insulation 1 3 1 600 1,800 Reinstallinsulation 2 3 1 600 3,600 BOILERMAKERS:
Weld preparation 1 2 1 600 1,200 EXAMINERS:
Perform exam 1 2 1 600 1.200 TOTAL: 54,600 i
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' Relief Reauest No. BV1-B3.120-2. Rev. O I I
- k remot'e visual examination from the inside of the pressunzer was considered as an alternative to the UT I L examination. A screen located at the surge line nozzle and baffle plates in the lower section of the l
} pressurizer would restrict access to the area ofinterest. The distance from the manway to the surge nozzle !
area is approximately 40 feet, making positioning adjustments of the remote camera difficult. Because of l these limitations, a remote visual examination is not considered a viable altemative.
i i l- Several methods are available to detect leakage from this area if a through-wall leak occurred. Listed )
l below are some examples:
8 i !
I a. The control room operators perform Operation Surveillance Test (OST) 1.6.2 " Reactor Coolant 1 j System Water Inventory Balance" every three days when the plant is opersing at steady 1 conditions. Leakage through the subject welds would be discovered by the conduct of this )
[ OST. j J
l
- b. Containment airborne radiation monitors continuously sample the containment atmosphere and
- alarm in the control room. The sensitivity of this detection method is dependent on the size of ;
the leak, reactor coolant activity, and containment background activity, j
! c. Leakage from this area would cause an increase in containment pressure, temperature and j
humidity which are indicated in the main control room. The containment pressure alarms in the j main control room. l i d. Substantial leakage from this area would collect in the containment sump. The sump level is l indicated and alarmed in the main control room.
! There are no credible failure mechanisms other than fatigue for this area. Corrosion degradation protection j is provided by the combination of the austenitic stainless steel cladding of the surge nozzle inner radius and i by the chemistry controls on the reactor coolant system. Strict chemistry standards are maintained to
- ensure a non-corrosive environment. Oxygen, chloride, fluoride and other contaminant concentrations are maintained below the thresholds known to be conducive to stress corrosion cracking. Since the surge nozzle is cast, the typical failure mechanisms associated with weld material do not apply to this examination. Erosion and Erosion / Corrosion degradation is not credible at this location. The austenitic stainless steel cladding resists this mechanism. There is relatively low fluid velocity in the surge nozzle and reactor coolant chemistry minimizes the amount of particles in the fluid thet could potentially cause erosion. Creep and stress relaxation are not concerns for the surge nozzle inner radius area since the design temperature of 680F is below the temperature where creep becomes a concern.
Fatigue degradation is a concern in this area due to the potential thermal cycling caused by the insurge and outsurge of the reactor coolant flow. Since the surge nozzle is cast with the bottom head, there is no nozzle to vessel weld. The inner radius is believed to be less susceptible to fatigue problems than a nozzle i to vessel weld. Initiation of fatigue cracking may have equal potential at the inner radius as compared to a nozzle to vessel weld, but the chances of having a pre-existing flaw are less likely in the 'mner radius casting than at a nozzle to vessel weld due to the manufacturing process This hypothetical flaw would then had to of been overlooked by the shop NDE (which includes surface, UT and RT examination). Inservice fatigue 1
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- Page 4 of 4 Relief Reauest No. BV1-B3. I20-2. Rev. 0 1
) crack gr'owth for such a flaw would be very small since the pressurizer is hot during the insurges and
] outsurges resulting in relatively high fracture toughness of the material. A thermal sleeve, installed in the l i surge nozzle provides a measure of protection from the affects of fluid temperature changes. Examinations !
are performed on the nozzle to safe end weld, which is within 18" of the inner radius. The nozzle to safe
{ end weld has bem satisfactorily examined without limitation during the first two intervals.
I The radiation exposure associated with the preparation activities for this examination is considered a
- significant hardship. If the preparation activities were performed, the subsequent examination would not 2
! significantly increase the level of quality and safety due to the low probability of the presence of s flaw in j this area based on the information presented above. It is therefore concluded that the intent of j 10CFR50.55a(a)(3)(ii)is met.
j ALTERNATIVE EXAMINATION
! A visual examination (VT-2) of this area will be performed in conjunction with the boric acid walkdown, i l performed every shutdown. Also, this area is included and documented in the Mode 3 walkdown of the j
- RCS boundary, performed during each startup following refueling outages as required by Item No. I B15.20. Both of these activities are performed by qualified VT-2 exammers. These examinations are l
} augmented by the leakage detection methods noted above. If the insulation is removed for maintenance or j other purposes, the UT exam of the inner radius section will be performed. l l
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