ML20002C522
| ML20002C522 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 06/22/1973 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML20002C517 | List: |
| References | |
| SR-14, NUDOCS 8101100409 | |
| Download: ML20002C522 (24) | |
Text
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1 CONSUMERS POWER CCMPANY Big Rock Point Nuclear Plant SPECIAL REPORT l
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Big Rock Point Plant Operations 1.
Reactor Water Chemistry-Fuel Crud Deposition 2.
Cycle 10 Fuel Performance 3
Cycle 11 Start-Up Report 1
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Prepared by: Consumers power Company June 22, it173 ppl/ 0o'/04 9
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I.
REACTOR WATER CHEMISTRY-FUEL CRUD DEPOSITION A.
INTRODUCTION Enough experience and data has been accumulated over the years to convince us that we now understand the full dimensions of the Big Rock Point reactor water chemistry-fuel crud deposition problem. Most important of all, we feel we have implemented satisfactory solutions.
However, only future fuel performance can confirm this judgment.
The sources of the impurities making up the fuel crud composition have been identified and have been eliminated (refer to Table 1). In addi-tion, the peculiar crud deposition pattern on the E-fuel has been remedied.
B.
BACKGROUND A review of the archives files on the Big Roch Point Plant show the reactor water chemistry and crud deposition problems to be the most persistent of the problems experienced at the plant over the years. The basic difficulty in defining and solving these problems has been related to our inability to get representative water samples for analysis from the t
primary and feed-water system and a good measure of the total amount of the crud deposition on the fuel. As a result of the recently completed extensive water chemistry program and fuel profilometry study, enough quantitative data are now available to draw some firm conclusions.
The first fuel inspections at Big Rock Point revealed crud on the fuel. Chemical analysis showed the crud to consist mainly of zinc, nickel, iron and copper - clearly constituents of the feed-water heater tube material. In March 1968, the feed-water heater tube material was changed from copper-nickel to stainless steel. Subsequent feed-water testing confirmed that this change effectively elidnated the main source of copper, zine and nickel in the feedwater.
During the short operations cycle (Cycle 5) following the change-t over of the feed-water heaters, the water chemistry-crud deposition ap-peared to be well under control, and Consumers Dower was confident that a significant corner had been turned even though Cycle 5 was of short dura-tion (four months) and with a peak reactor power of only 60 MWe.
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Cycle 6 (July 1%8 to May 1%9) started off at 75 Mwe reactor power but because of high off-gas the load was reduced partway through the cycle to 60 MWe. It was also noted during this cycle that the reactor water pH varied considerably. This indicated that the primary system was seeking a new chemical equilibrium because of the changeover of copper-nickel feed-water tubes to stainless steel. At the end of Cycle 6, crud deposition measurements clearly showed that the situation was worse than ever. Additionally, the crud compo:ition had changed to predominantly copper from the earlier zinc, nickel, copper composition.
An extensive cooperative water chemistry study program with General Electric Company was instituted for Cycle 7 (May 1969 to February 1970). - This study showed that the feedwater had been cleaned up.
There was sone evidence that small amounts of impurities were injected from the demineralizers to the primary system after " fluffing" of the beds, shut-r downsand/orregenerationoftheresinsbutthiscouldnotaccountforthe levels of copper seen in the reactor water. The conclusion was that the i
source of copper had to be in the primary system. The only pocsible sources were " hideout" on the surfaces of the primary system including the steam drum and the copper alley tubes in the clean-up system heat exchangers (both regenerative and nonregenerativa).
The reactor water chemistry appeared to stabilize during Cycle 7 Crud deposition measurements at the end of Cycle 7 showed an improvement in the deposition rates. This improvement continued through Cycle 8 (March 1970 to February 1971). The total amount of crud deposited on fuel rods dropped significantly from Cycle 6 to Cycle 7, then again from Cycle 7 to Cycle 8.
However, the peak crudded area of the fuel (the lower quarter) still exhibited the same peak rod crud deposition rate.
During Cycle 7, a series of tests was run on water entering and leaving the clean-up system heat exchangers. These tests revealed that the water was picking up copper from the copper alloy tubes in the clean-up system heat exchangers. Profilometry studies on the crud deposition patterns in fuel assemblies showed that crud was preferentially being t
deposited on the lower quarter of the fuel assemblies on the outer row
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of fuel rods facing the channel walls. Knowing the crud thickness and i-(
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3 this pattern made it possible by using mass balance techniques to conclude that most of the copper in the crud deposited on the fuel could be attrib-uted to the copper picked up in the clean-up system heat exchangers.
At the end of Cycle 7, the internal surfaces of the steam drum were inspected. The inspection proved the steam drum to be relatively clean with no large visible accumulations of ciud. Considering the purity of the feedwater, it was concluded that the clean-up system heat exchangers constituted the major source of copper for the crud deposition.
Also, during Cycle 8, it was determined that there was a signifi-cant bypass flow (approximately 13 gpm) between the tube and shell sides of the clean-up. system regenerative heat exchangers. Previously, during Cycle 7, a small leak rate had been calculated (approximately 2 gpm).
This increase in bypass flow further confirmed that the copper-nickel tube clean-up system regenerative heat exchangers were a source of copper in the reactor water.
C.
PROBLDi c
Only in mid-1971, after all the above experience and data had been accumulated, could the full dimensions of the reactor water chemistry-crud deposition problem be appreciated.
First, the source had to be identified, but it chang,ed with time.
Initially, it was the feed-water heaters; then it was " hideout" or just an inventory of feed-water heater tube material in the primary system. _ These sources masked two other lower level sources that probably contributed im-purities from " day one."
These other sources were the clean-up system heat exchangers and the impurities washed off the condensate demineralizers by " fluffing," after regeneration or after a shutdown.
Secondly, some explanations for the crud deposition pattern on the fuel were to be provided to help explain the cause of the fuel failures.
Initially, on the B-and C-fuel, there was so much crud available it de-posited fairly uniformly over the fuel.
(Crud depositions were not as limiting on B-and C.= cype fuels as they are on E-and F-type fuels because of greater heat transfer areas associated with the B-and C-fuels.) Then, as the primary system cleaned up a bit after the feed-water heater change-
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over to stainless, the crtd began to deposit preferentially in lower i
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positions yet higher power regions of the fuel rods (exterior fuel bundle rods). The flow tests also explained in part the preferential accu =ula-tion of crud on the lower quarter of the outer fuel rods.
D. ' CORRECTIVE ACTION The solution to the reactor water chemistry-crud deposition problem consisted basically of eliminating the source of the impurities that made up the crud. The following have been done:
1.
The feed-water heater tube bundles were changed to stainless steel in March 1%8.
2.
Operating practices with the condensate demineralizers were reviewed and modified. After May 1970, the resins were no longer regen-erated. This effectively eliminated the " spikes" noted in the water chemistry study program.
3 The clean-up system heat exchangers were replaced in April 1][2.
In addition to the above, the original fuel channel-orifice hardware on 69 of the 84 fuel support-tube-and-channel e.ssembliet has been replaced during the refueling outages in 1972 and 1973 The modified support-tube-and-channel assemblies improve the flow characteristics in the lower quadrant of the fuel bundle.
E.
RESULTS Fuel performance during Cycle 10 (May 1972 - February 1973) in-proved markedly. More heat was produced with smaller off-gas releases than in the several previous cycles. Continued improvement in fuel per-formance is expected as copper " hideout" sources are depleted and existing fuel that has been exposed to previous water chemistry conditions is re-placed by new fuel.
II.
CYCLE 10 FUEL PERFORMANCE (MAY 1CJ72 - FEBRUARY 1973)
A.
INTRODUCTION Cycle 10 fuel examination was conducted during the March 1973 refueling oui, age. The primary purposes cf the irradiated fuel inspection 4
were to characterize Cycle 10 fuel performance, obtain information con-cerning crud deposition and to enable the respective fuel supplier to col-lect data to verify several of their design models such as-fuel rod growth, I
clad creepdown and pellet stack shortening.
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Consumers Power Ccanpany, General Electric Company, Exxon Nuclear Company and Battelle Northwest Laboratories participated in the. fuel exte::1-nation. The inspection activities were perfomed by two independent teams.
The first consisted of General Electric personnel and the second was com-prised of indivic'uals from Exxon Nuclear and Battelle Northwest. Consumers Power Company's participation in the examination consisted of performing dry sipping tests on all assemblies scheduled to be returned to the core; scheduling and coordinating the examination activities; assisting the in-spection teams in perfoming specific tests; and, analysis of selected i
portions of the data.
B.
WORK SCOPE The inspections performed consisted of visual examinations of exterior surfaces of selected fuel assemblies and fuel rods; profilometry; j
gamma scanning; eddy-current testing; and, fuel rod length measurements.
l The specific work scopes are detailed below:
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a.
Visual Examination, General Electric Company and Consumers Power Company per-formed a visual examination of ten fuel assemblies containing individual 4
rod failures to determine the failure mechanisms. The fuel assemblies in-cluded in the examination were four (4) Type F, three (3) EEI mixed-oxide and three (3) modified EG assemblies. The exposure range for these assen-blies was 6,300-15,700 mwd /T. A visual examination of twelve (12) remov-able EEI mixed-oxide rods was also perfomed.
b.
Profilometry Analysis Twelve (12) fuel rods and fourteen (14) nonfueled rods (cobalt target) from F-type assemblies were removed and profiled to obtain information concerning crud thickness, clad creepdown and clad ovality.
c.
Length Determination Length measurements were performed on the fourteen (14) nonfueled rods to determine Zircaloy clad growth.
d.
Mixed-Oxide Fuel Examination The following examinations were performed by General
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Electric on mixed-oxide fuel:
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(1) Visual examination of the three EEI mixed-oxide fuel assemblies and twelve EEI mixed-oxide removable rods.
(2) Sipping tests were performed on eight of the EEI removabln rods to determine cladding integrity. In an attempt to locate the failed rod (s) in assembly EP-02, all eight gasoline-uranium oxide rods were removed and the bundle was resipped. The resulting sip signal showed the assembly to be failed, thus indicating a failed Pu rod.
(3) Two mixed-oxide rods from assembly EP-02 were profiled to obtain data on crud thickness.
- 2.. Exxon Nuclear Exxon's fuel inspection focused on three (3) assemblies (D-70,
-71, and -72) and a total of twenty (20) individual fuel rods withdrawn from these assemblies. In addition to the examinations described in Para-graphs a and b under " General Electric," Exxon perfomed the following tests on the twenty (20) withdrawn' rods:
a.
The fuel rod lengths were measured to determine the amount
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of rod growth, b.
Eddy-current and ultrasonic testing were performed to locate and define cladding defects.
c.
The rods were gamma-scanned to detect changes in the pellet stack length and to verify that there were no significant gaps in the pel-let stack, d.
Of the 20 fuel rods inspected, two were mixed-oxide rods.
These two rods underwent the same examinations as the uranium rods.
3 Results of the Cycle 10 Fuel Examination a.
Fuel Failure The sipping results revealed that 23 of 84 fuel assemblies contained failed rods. The end-of-life exposures of 17 of the failed assembliesexceeded10,000 mwd /T. Visual examination of the failed assem-blies revealed the probable cause of the failures to be accelerated corro-l sion. Evidence of internal hydriding was not observed. Table 1 provides a detailed description of the Cycle 10 core composition and identifies the assemblies that failed.
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b.
Mixed-Oxide Fuel Performance The three (3) mixed-oxide fuel assemblies (EP-01, -02 and
-03) also contained failed rods. Their bundle exposures were approximately 15,400 Wd/T,15,700 Wd/T and 15,000 Wd/T, respectively.
Assembly EP-01 was visually inspected in an attempt to determine the cause of failure. Only one rod exhibited slight crud spalling (located in the middle of the lower tier). All the other sides of the assembly appeared normal. Visual examination of the peripheral rods did not reveal the cause of the failure.
Assembly EP-02 likewise was visually inspected. Slight crud spalling was observed on one exterior rod. All other sides of the assembly appeared normal. A further attempt was made to determine the location of the failed rod (s). All eight (8) gadolinia-uranium oxide rods were removed and the assembly was resipped. The resulting sip signal showed the assembly to be failed, indicating a Pu rod failure. In addi-tion to the above inspections, tro exterior rods were profiled to deter-i mine crud thickness. The crud thickness noted is equivalent to that found on other 3-cycle fuel assemblies.
Fuel assembly EP-03 was reconstituted using EP-01 as a source of acceptable fuel rods. The' fuel rods removed from EP-03 exhibited crud spalling and white zirconium oxide. These observations indicate that the rod failures were due to accelerated corrosion of the cladding from the outer diameter surface. This type of failure mechanism has been noted on uranium oxide fueled rods and is presently the predominant cause of fuel failures at Big Rock Point.
Twelve (12) EEI mixed-oxide rods from various uranium fueled assemblies were also examined. Two of the rods were withheld from future operation since clad corrosion attack was observed. These two rods were not sipped. The exposure for these two rods was approximately 17,000 Wd/ Teach. The remaining ten rods were found to be sound. General Electric selected four of the ten rods for shipment to UNC for post-irradiation examination. The remaining six rods were returned to the core for further irradiation.
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Two mixed-oxide rods from assembly D-72 were examined by Exxon Nuclear. The scope of the inspection is given in Section II, Paragraph B.
Both of these rods were sound.
c.
Crud Deposition A comparison between profilometry data for Cycles 9 and 10 revealed that the peak diametrical crud buildup has decreased from approxi-mately 0.006 inch to 0.0045 inch. An increase in loose flaky crud was observed for Cycle 10, indicating that the crud morpholoEy is being altered to a more permeable and less tenacious nature, d.
Effect of Redesigned Support-Tube-and-Channel Assemblies In 1972, forty (40) redesigned support-tube-and-channel assemblies were inserted in the core. Their effect on fuel performance cannot be evaluated at this time because of insufficient information.
Higher exposures are required on fuel assemblier that have been restricted to operaticn in the new channels before a judgment can be made.
4.
Conclusion Since -it has been shown that the presence of deposited crud on '
fuel rods is contributing to excessive cladding colrosion, decreases in deposited crud, changes in crud morphology and improvements in feed-water impurity conditions are significant. These effects should provide for improved fuel perfomance during Cycle 11. Analysis of Cycle 10 fuel per-formance indicates the absence of failures due to internal hydriding of zirconium.
Analysis of examination data indicates that mixed-oxide rods fail from a mechanism identical to that causing uranium fuel rod failures.
When these failures have occurred, no plutonium has been detected outside of the fuel rods.
With known defective mixed-oxide fuel in the Big Rock Point reactor, we have been unable to identify any plutonium in the pri-mary coolant, liquid or gaseous effluent. Plutonium would be most readily detected through alpha activity. Measurement of gross alpha radioactivity is conducted on all effluents. If one made the very unrealistic assump-tion that every alpha particle detected came from plutonium-239, the i
maximum liquid effluent concentrations during the past several y irs I
would be:
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MaximumAlphaActivity-2 cpm /ml This is equivalent to approximately 2.7 x 10'6 pCi/ml in the waste tank
-10 or approximately 5 x 10 pCi/ml in the discharge canal. The permissible
-5 public drinking water limit for plutonium-239 is 5 x 10 pCi/ml.
Selected rods from the modified EG, high burnup Reload C, the corner rod experiment carrier assemblies, and four (4) EEI mixed-oxide rods were shipped to Vallecitos for possible postirradiation destructive examination. General Electric intends to obtain up to seven (7) additional mixed-oxide rods for possible postirradiation examination.
Future examination (Cycle 11) will center on the 11 x 11 fuel. This fuel has greater thermal margin (lower average heat flux) than our present 9 x 9 fuel and will significantly decrease problems due to accelerated cladding corrosion.
III. CYCLE 11 START-UP REPORT A.
INTRODUCTION Big Rock Point Cycle 11 was designed to operate between 200 and 220 MWT for approximately one year. The core loading includes eight (8)
General Electric Reload EG assemblies; sixty-five (65) General Electric Reload F assemblies; two (2) Exxon Nuclear J1 assemblies; two (2) Exxon Nuclear J2 assemblies; one (1) EEI-General Electric EP assembly; four (h)
Nuclear Fuel Services Demonstration Assemblies (NFS-DA); and two (2) Exxon Nuclear prototype Reload G assemblies. The prototype Reload G and the NFS-DA are new fuel types being irradiated for the first time at Big Rock 1
Point. Both are 11 x 11 array designs while the remainder of the core consists of 9 x 9 array fuel types that have been used previously. The J2, EEI, NFS-DA and G assemblies all incorporate mixed-oxide fuel in their designs.
B.
REACTOR START-UP Upon completion of core reconstitution, core shutdown margin verification was successfully performed as required by Technical Speci-fications Section 5.2.2(b). Each step of the verification was comprised of the complete withdrawal of one strong peripheral control rod plus an adjacent control rod withdrawn six notches. Computer calculations showed
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theextrasixnotchestobeworthinexcessof0.4%ak/k. Additional I
shutdown margin tests were conducted by entirely withdrawing two adjacent
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m TABLE 1 End of Cycle 10 Core Composition s
Bundle Exp No of Bundle Exp No of Bundle Exp-No of Type mwd /T Cycles Status Type mwd /T Cycles Status Type mwd /T Cycles Status F 6,654 2
F F-29 3,477 1
D-72 5,646 1
F-02 10,418 2
-F F-30 4,848 1
D-73 5,598 1
F-03 10,708 2
F-31 4,997 1
EP-01 15,416 3
F F-04 10,018 2
F-32 4,397 1
EP-02 15,712 3
F F-05 8,986 2
F-33 1,951 1
EP-03 15,026 3
F F-06 2,251 1
F-34 2,766 1
D-61 12,333 3
F F-07 10,043 2
F-35 5,073 1
D 62 12,418 3
F F-08 9,770 2
F-36 4,115 1
D-63 11,837 3
F F-09 9,606 2
F-37 5,267 1
E-55 7,767 2
F-10 9,964 2
F-38 5,436 1
E-61 16,390 4
F F-11 10,148 2
F-39 2,794 1
E-62 16,662
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F-12 6,289 2
F F-40 2,839 1
E-65 11,579 3
a F-13 9,430 2
F 41 5,448 1
E-66 11,635 3
F F-14 8,757 2
F-42 5,244 1
E-67 11,956 3
F F-15 10,833 2
F 43 4,042 1
E 68 11,927 3
F-16 9,946 2
F-44 5,014 1
E-70 7,346 2
F F-17 9,694 2
F-45 2,711 1
E-71 11,667 3
F F-18 8,748 2
F F 46 1,980 1
E-72 12,750 3
F-19 6,547 2
F F 47 4,470 1
E-74 14,360 3
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F-20 9,578 2
F 48 5,041 1
E-75 12,669 3
F F-21 9,222 2
F 49 4,832 1
E-78 12,507 3
.r-22 10,346 2
F-50 2,254 1
E-79 13,769 3
F-23 9,636 2
F-51 4,532 1
E-80 13,003 3
F-24 6,289 2
F F-52 2,263 1
E-81 14,067 3
F F-25 9,032 2
F-53 1,990 1
E-82 12,807 3
F-26 3,420 1
F-54 1,974 1
E-83 12,669 3
F F-27 2,217 1
D-70 10,210 2
E-84 13,966 3
F F-28 4,468 1
D-71 10,675 2
E-85 12,388 3
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strong peripheral control rods plus withdrawal of six notches of a neigh-boring control rod. As predicted oy computer calculations, the core re-mained suberitical during each step of the shutdown margin test. Final calculations indicated a total shutdown margin of 6% Ak/k-with the most valuable control rod (3% Ak/k) withdrawn from the core.
The first Cycle 11 beginning-of-cycle (BOC) cold critical con-trol rod pattern differed from the computer predication by one notch
(%0.06fAk/k).' Figures 1and2arediagramsofthepredictedandactual Cycle 11' BOC critical control rod patterns.
After completion of the first critical approach, the moderator temperature coefficient test was conducted (Technical Specifications Section 5.2.1).
Results of the test indicate a maximum addition of 12 4
cents from ambient (s70 F) to 137 F, well within the Technical Speci-fications limit of one dollar. Figures 11 and 12 are plots of p vs temperature and the temperature coefficient vs temperature (p/T).
Fluxwires were irradiated upon reaching equilibrium conditions at a power level of 216 MWT. Figures 3 through 10 are comparisons of the actual vs computer-predicted axial flux distribution for the eight (8) in-core monitor locations. These figures indicate a very good match between the predicted and measured axial flux distributions.
In summary, the physics start-up was uneventful. All predicted and measured values were in good agreement and well within Technical Specifications.
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Figure 1 Predicted BOC 11 Cold Critical 1
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J ENCLOSURE SAMPLE TECIMICAI, SPECIFICATION LANGUAGE The following language should be substituted, as appropriate, into the Technical Specifications 'where existing surveillance requirencnts are superseded by ASME Section XI inservice inspection and testing requirenents:
Inservice inspection of ASME Code Class 1, Class 2 and Class 3 a.
components shall be perforned in accordance with Section XI of the ASME Boiler and Presure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(1).
b.
Inservice testing of ASME Code Class 1, Class 2 and Class 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except
^
where specific written relief hes been granted by the NRC pursusr.t
- o 10 CFR 50, Section 50.55a(g)(6)(i).
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PART 50 o LICENSING OF PRODUCTION AND UTILIZATION FACILITIES the preservice examination requirtrr.ents 37 b) *For construction per issued Code for Pumps and Valves for Nus.r set forth in Eection XI of ed.tions of the or af ter July 1.1974, pumps which Power and Addenda
- and the require-ASME Boiler,and Pressure Vasel Code he part of the reactor coolant pressurc ments appheatle to valves set forth in and Addenda applied to the :onstruc-boundary
- shall meet the requirements articles 1 and 8 of editions of section III tion of the particular component fn ac-for Chss 1 components set forth in Sec-f; of the ASME Boller and Pressure Tessej cordance with paragraph (c). (dt. 'e t,
tion III of the ASME Doller and Pres-g Code and Addenda or for Class 1 valves or f f s of this section.
sure Vessel Code and Addenda" in ef.- of section III of the ASME Boiler and alli Components whMh are clay 1fied Pressure Vessel Code and Addenda in as ASME Code Class 2 and Class, rmd e
R fect' on the date of order' of the cump[c: eCec} 12 months prior to the date of is-supports for components which are cla';;
or 12 months prior to the formal dockctn suance of the construction permit is re-
" date of the at' plication for construction quired. The vahes may meet the. re-sifled as ASME Code Class 1. Class 2. M7 cuirements set forth in editions of these Class 3 shall be des ened and be pros:ch.
permit. whichever is later: Prorided' with access to enable tie pcrio*.mance o That the applitable ASME Code prost-Codes cr Addenda which have become tions for pumps shall be no earher than effective af ter the date of valve order or inservice examination of suc.
00...+
af ter 12 months prior to the date of nents and shall meet the preservice cx-epn a
ett amination requirements set f orth in Scc-191 editio.
forth in subsecuent issuance of the construction permit.
tion XI of editions of the ASME BQr reoutrements set editions of this Code and Addenda which (36 For construction permits issued on and Pressure vessel Code and Addcn-become effective.
or after July 1.1974. valves which are da " apphed to the ecnstruction of t.n part of the reactor coolant pressure particular component.
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gg y,7y,3*
boundary shall meet the requirements (1111 Pumps and valves which are clas-8 (1) For construction permits issued for Class 1 components set fotth in Sec-sided as ASME Ccde Class 1 shal: bc dc-before January 1.1971. for reactors not tion 111 of the ASME Boiler and Pres-signed and be provided with acccn t?
licensed for operation. valves which are sure Vessel Code and Addenda " in ef-enable the performance of insenice tot-en N M of der' M M vaM ing M h pum M Tdm M N; M dary shal eet t!
equie en or 12 months prior to the formal docket operational readme% set fcrth in Mr-se rW in date of the application for construction tion XI of editions of the ASME Bohr (D The American Standard Code for Perrnit, whicheser is later: Protided, and Pressure Vesstl Code and Adde:a Pressure Piping (ASA B31.1). Addenda.
That the applicable ASME Code provi-applied to the ccnstruct:cn of the pr-sions for valves shall be no earlier than ticular pump or valve m a:cordancey S an de r Pre sure 1 in" those of the Winter 1972 Addenda of the paragraphs tei and < f
- cf thh n.
US d
1971 edition. The valves may meet the or the Summer 1913 Addenda.whichncr cab! C e Ca. n effe t o he ate rcquirements set forth in subsequent is later.
of order
- of the va!ves or the Class I editions cf this Code and Addenda which sive Pumps and valves which arc (,.c':
sined as ASME Code Class 2 and C171 section of the Draf t ASME Cody for b(come ef'ective.
a Pumps r.nd Vahes for Nuclear Power /
(g> Inservice intrection requirements:
shall be desizned and be pro'.tdec. 7 m Addenda, and Code Cases in eff ect on the (1) For a facihty whose construction m cess to enabit the performance (..
date of order of the vahn;or permit veas issucd prior to January 1. $ senice testing of the pumm and '.
w-(ii) The nondestructive examination 1971. comoonents imeludms: sunporto 3 f or assessme opctationa! rradu ' s F and acceptance standards of ASA B31.1 shall meet the requir(ments of para-c forth in Stetten XI of d2n* Of S graphs eg e '4 8 and m i$> of this section " Boiler and Pre =sure Vessel Ccr tE onrd fr ta the extent practical, Cumpunente g Addenda htteaep e
which are part of, the mictor (oulant of the particular '""n!' o' u!v#u hichemlr t y*
(
T valves set fmth in the Draft ASMI,'
Code for Pumps and Valves for Nuclear pre"ure bour.dary and their simrorts Summer 1973 Acccnda.
Power and Addenda in clicct on the date shall mcot the requirements appheable later.
of order of the talves may be apphed.
to components which are classified as tv' All components (includinr sa
- g The valves may meet the requirement % $ ASME Code Class 1 Other safety-re-portsi may meet the recunemenb *e set forth in editions of ASA E31.1. USAS r* lated pressure vescels. pipmt pumns and forth in subecqucnt editrns of ce s c.
[ B31.10. and the Draf t ASME Code for $ valves shall meet the reouirernents ap.
and addenda or part:on= thereci whr.
Pumps and Valves for Nudear Power. ; p.icable to com'somnts which are classi-become effectis e.
g Addenda, and Code Cases, which became
- fled as ASME Code Class 2 or Class 3 e41 Throurhcut the service life of a effective af ter the date of order of the '
'2n For a faciutv uhose construction faci!!ty, componc nts sincludmc sup vah cs.
permit was l'. sued on or af tcr January 1.
ports > which are classined as ASME Ccu (2) for comtruction permits iwued on or 1971. but before July 1.1974. component ^
Class 1. Class 2 and Class 3 shall mW
. tsus bdore Juh L sincludinc supports) uhich are clasa-the requirements, except dairn anu : -
va es t I f ed as ASME Code Class 1 and Class 2 cess provisions and prescry:ce cr.mma-197
- ate hart of the reactor coolant pressure shall be designed and be provided with tien requirements. ret fcrth m Se;"
boundary 'shall mect the requirements access to enable the perfor: nance of it' XI of edit: cts of the ASME' U2Mer am for Class I valves set forth in editions c; inservice examination of such compo-Pressure Vessel Code and Addtnda (il the Draft ASME Code for Pumps and Valves for Nuclear Power and Adden.sa' nents asneludin;; supports and si1> tests that become eJecthe subsequent to (O in effect
- cn the drte of order' of the for operational r(ad. ness of pumps and tions specified in pigraphs E * *2' O.n valves and the requirernen;s applicaNe to valves, and shall meet the preservice ex-
.g.(3i of this secuon and are mcorp valves set forth in articler 1 erd 8 of amination requ:rements set forth in edi-rated by reference in par.! graph
- b>
c editions of section 21I of the ASME BoPer 1:ons of Section XI of the ASME Boller thw section. to the cuent tractt al wC and Pressure Vessel Code and Addenda
- and Pressure VesselCode and Addenda in the limitations of des:;n. recmctr in effect on the date of order of the in et'ect 6 months prm to the data of and materials of construction of th valves, or (ID the renmrements apph.
Issuance of the construction pcrmit. Tne components.
cable to Class 1 valves of section III of components ' including supports > may sti The initial insenice examination the ASME Boiler and Pressure Vessel meet the requirements sct forth in sub-conducted during the first 40 men:h Code and Addenda in e:!ect on the date sequent editions of thh code and ad-shan comply with the reauiremen" 1 of order of the valve; Protided. hotrerer.
denda uhich become c:!ective.
the editicas of the code and addend; ;
That it the valves arc ordered more than i36 For a fatihty whose construction ef!cet no incre tnan G mcnths puor t 12 months prior to the date of hsuance of permit was issued on nr af ter July 1 the date of start c! facihty commeren the construction permit, comphance with 1974:
operation.
the requirements for Class I valves set (16 Components which are classined in, The inservice examinations con l forth in editions of the Draf t ASME as ASME Code Class I shall be designed ducted during successive 40-month pt and be provided with access to enable s throughout the service lite of th see page 50 is for footnotes i enrough 6.
the performance of inservice examina-riop!ity thereaf tcr shall comp *> w:
fact
- AmenAJ as t it uso.
tion of such cc:nronents at.d shall mect P00R ORIGINf"~
PART 50 o LICENSING OF PRODUCTION AND UTILIZATION FACILITIES those requirements in editioC
-'f the the licenste that could result if I
code *and addenda in effect no. a than requirements u are imposed on J
, C Nont's prior to the start of each 40-facility.
n n.
th period.
(ID The Commission may require the
- (iii) The initial inservice tests of licensee to follow an augmented inservice pumps and valves for assessing opera-Impection program for systems and com-tional readiness and system pressure ponents for which the Commission deems tests conducted during the first 20 that added assurance of structural rell-months shall comp!y with those require em abuity is necessary.
ments in editions of the code and adden. U thi Protection systems: For construc-da in e!!ect nc more than G months prior cc tion permits issued af ter January 1.1971.
to the start of facility commercial opeaa
- protection systems shall meet the re-tion.
Equircments set forth in editions or revi-(iv) Inservice tests of pumps. and sions of the Institute of Electrical and valves for assessing operational readi-Electronics Engineers Standard: "Crl-ness and system pressure tests conducted teria for Protection Systems for Nuclear during successive 20-month periods Power G(nerating Stations." #IEEE-T9) thrcughout the service life of the facility in e!!cet' on the formal docket date of shall comply with those requirements in the application for a construction per-editions of the code and acidenda in ef-mit. Protection systems may meet the fect no more than 6 months prior to the requirements set forth in subsequent edi-start of each 20-month period.
tions or revisions of IEEE-279 which tv) For an operatmg facihty whose become e!!ective.
operating liecnse was iv.ued prior to
~
March 1.1976, the provisions of para- -
graph (g i a 41 of this section shall become e!!cctive af ter September 1.1976, at the start of the next regular 40-month pe-riod of a series of such periods beginnmg at the start of f acility commercial opera-tion.
($1 (1) The inservice inspection pro-gram for a facihty shall be revhed by the licensee, as necessary, to meet the requirements of paragraph (g) t4) of this section.
t iin If a revhed inservice inspection eR _ program for a facility conf'.fcts with the o technical specification for the facility.
[ the licensee shall apply to the Commis-sion for amendn.ent of the technical v specif.cationa to wnictm the tecnmcal specification to the nevaec'.,rocrun.This application shall be sul?tted at least G months before the start of the period during which the proitsions become ap-pHeabic as detcrmir.ed by paragraph (g>
(4 8 of this section.
t ili p If the hecmcc has determined that confortnance with certain code re-quiremsnts b impractical for n!$ f acility, the lleensee shall no:1fy the Commission and submit information to support his determinations.
(iv) Where an examination or test reautrement by the code or addenda is determined to be imprac*ical by the 11-censee and is not include n the revhed inservke inspection program as permit-ted by paragraph 8g'i4 of this section.
the basis for this deterraination shall be demonstrated to the satisfauion of the CommisMon not la ter than 12 months after the exmration of the initirl 120-month period of operation from start cf facility commercial operation and each subsequent 120-month period of opera-tion durmg which th2 examination or test is determined to be impract: cal.
(6)(D The Commt=sion will evaluate determinatiens under paragraph 'g)(5) of thh section that code requirements are impractical and may grant such re-lief as it determines is authorized by law and will not endanger hfe or property or the common defense and security and is otherwis-in the pubtle interest givmg
[
1 t
due conside.ation to the burd(n upon February 27,1976 50-14b P00RORIBIL'.I1
_t....-
.