ML20005G816
| ML20005G816 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 01/16/1990 |
| From: | Long R GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TASK-02-01.B, TASK-02-01.C, TASK-03-05.B, TASK-03-06, TASK-05-06, TASK-2-1.B, TASK-2-1.C, TASK-3-5.B, TASK-3-6, TASK-5-6, TASK-RR NUDOCS 9001230132 | |
| Download: ML20005G816 (17) | |
Text
{{#Wiki_filter:.- e OPU Nuclear Corporation Sr One Upper Pond Road i Parsippany, New Jersey 07054 201-316-7000 ) . TELEX 136-482 Writer's Direct Dial Number: January 16, 1990 i l U.S. Nuclear Regulatory Commission Attention: Document Control Desk Mail Station Pl-137 Washington, D. C. 20555 I Gentlemen: i Subjects Oyster Creek Nuclear Generating Station Docket No. 50-219 N.J. - DEP comments on Draft FTOL SER We have received your letter dated Dec. 22, 1989 which forwarded concerns l oxpressed by the Stato of New Jersey, Department of Environmental Protection (DEP) regarding several technical issues contained in the Draft Safety ? Evaluation Report (SER) for the Full Term Operating License (FTOL) for the Oyster Creek Nuclear Generating Station (OCNGS). Your letter requested GPU Nuclear (GPUN) positions on these issues. The NJ-DEP letter dated December 11, 1989 consists of the following sections: History i Overview of the SEP for OCNGS I. Overall Review (Independent Contractors) II. BNE In-house Review III. BNE SEP Topic Comments i l We have reviewed each section and our assessments are provided below. + t Historv/ Overview of the SEP for OCNGS t The contents of these t.wo sections are excerpts from NUREG-0822 " Integrated Plant Safety Assessment Report (IPSAR)/ Systematic Evaluation Program (SEP)/ Oyster Creek Nuclear Generating Station" which was published by the ~ NRC in January, 1983. They provide historical descriptions'of Oyster creek j licensing proceedings and SEP review. 1 9001230132 900116 C1 ) PDR ADO ^K 05000219 P PDC L J- / x hk } C320467 i p\\ N. GPU Nuclear Corporation is a subsidiary of General Pubhc Utihties Corporation ,/
/. pverall Review (Independent contractors) According to the NJ-DEP letter, the Nuclear Engineering Section (NES) of the Bureau of Radiation Protection hired an independent contractor, Mr. Peter Davis, in 1985 to review the oyster Creek IPSAR. The NES hired Mr. Davis because of his experience in the application of Probabilistic Risk l Assessment. During the oyster Creek SEP review by NRC, 40 topics were identified, after j the screening process described in IPSAR, as having certain aspects of plant design differing from current criteria. These topics were considered in the integrated assessment of the plant, which consisted of evaluating j the safety significa.nce and other factors of the identified differences 'l from current design to arrive at decisions on whether backfitting was necessary from an overall plant safety viewpoint. To arrive at these decisions, the NRC staff relied on their engineering judgement coupled with the extensive experience base developed by the NRC and/or deterministic analysis. In order to supplement the staff's judgement concerning safety i significance, a Probabilistic Risk Assessment (PRA) was performed by the NRC contractor, Sandia National Laboratories, to the extent possible. For reasons given in Appendix D of IPSAR, only certain topics could be readily analyzed by a PRA. Of a total number of 40 topics considered in the 7 integrated assessment, 20 were evaluated using PRA techniques. As documented in the NJ-DEP letter, Mr. Davis concluded that "The risk assessment evaluation of 20 SEP topics contains several apparent discrepancies. In no case, however, were such discrepancies found to have i an adverse influence on final decisions regarding backfit requirements. The qualitative results used for backfit decisions are, therefore, considered valid." l l The following GPUN responses are provided for each of the BNE concerns raised in this Section. BNE CONCERN "For a large number of deleted SEP topics, the NRC evaluation does not provide the basis on which they were deleted for the OCNGS". GPUN RESPONSE: The deletion of topics to arrive at the 40 topics considered for backfit in the integrated assessment was conducted by the NRC,.and CPUN cannot authoritatively comment on the basis. However, the process is described in Section 2 (pages 2-1 through 2-4) and outlined in Table 2.1 (page 2-4) of IPSAR (NUREG 0822). Appendices A, B, C, E and Section 3 briefly describe, with references, the basis by which the 137 topics on the final list of Phase I topics reviewed during Phase II, were reduced to the 40 topics considered for backfit. The basis is sometimes provided in very brief summary form, without the details that a reviewer unfamiliar with the topic would require in order to independently review and confirm the NRC basis for deletion. This does not reflect a deficiency in the NRC analysis, but merely the reality of the difficulty of documenting a long complex process in a one volume report. e C320467 ~ 1
t HME_CQHCIEN1
- The quantitative probabilistic risk evaluation of SEP, topics was a very minor consideration in deriving backfitting requirements.
Instead, it l appears that the primary basis for decisions on backfit requirements was engineering judgement coupled with the extensive experience base developed by the NRC in using design basis accident concept in conjunction with deterministic analysis." GPUN RESPONSES: We agree that the PRA was not a major consideration in deriving the backfit requirements, which were primarily based on engineering judgement, experience and/or deterministic evaluation. The results of the PRA were utilized for assigning the priority (i.e., schedule) for the backfit requirements. We believe this approach is conservaties and appropriate. BNE CONCERNt "The risk assessment evaluation of 20 SEP topics as provided in the Attachment Section, contains several apparent discrepancies. In no case, i however, were such discrepancies found to have an adverse influence on i final decisions regarding backfit requirements. The qualitative results i used for backfit decisions are, therefore, considered valid". 5 GPUN RESPONSES.L l We agree that while there are undoubtedly some discrepancies in the risk assessments presented in the IPSAR Appendix D, they do not invalidate the results and conclusions. BNE CONCERN: "Several topics were identified wnich appear to have an influence on the probability of causing releases below levels considered in PRAs, but in i excess of Protective Action Guides." GPUN RESPONSE: The contractor study conducted for the BNE, in addition to reviewing and commenting on the IPSAR, reviewed the issues to identify those which might t have an impact on the probability of offsite releases less significant than those considered in PRAs (which focus on core melt) but'in excess of EPA protective action guides (PAG). Several topics were identified. While this may be of some interest, the contractor correctly states that such events are not significant contributors to public health risk. It should l be noted that the EPA PAG limits are extremely low when compared to 10 CFR 100 siting criteria. For example, a whole body PAG limit is 1000 mrem i while a 10 CFR 100 siting criteria limit for whole body dose is 25,000 mrem y for the exclusion area and low population zone. The PAQ limits are not intended to be used to establish siting or licensing decisions. However, to exceed even the relatively low PAG dose levels offsite,. multiple safety systems must fail and very severe core damage must take place in addition i to adverse meteorology. C320467 k
Th3 BNE rCport Ctct3D that th y did ntt calculete "th3 amount of radioactivity release required to exceed PAG guides".. However, using the-unusumptions made by BNE (i.e.,rbased on the reactor coolant system (RCS) inventory without core damage) we conclude thst the RCS simply-does not-contain enough source term to reach the Protective Action Guide limits offsite. BNE CONCERN "The Oyster Creek core melt probability (4.4E-4 per.Section IIA) as estimated from revising the results of the Millstone Point Unit'l PRA is-relatively high,. as is the frequency of major release probability.when compared to other PRA results. These results should be viewed with' caution, however, due to the inherent uncertainties in PRA results as well as variations in rigor and methodology employed by the PRA studies and uncertainties in extrapolating the Millstone results to Oyster Creek. An estimate of risk from Oyster Creek due to' core melt accidents.is provided in the Attachment Section." GPUN RESPONSE 1 s The BNE contractor estimated a core' melt probability for Oyster Creek (and -i a probability of major releases) baoed on a superficial comparison of. 4 Oyster Creek with Millstone Point Unit 1 and revision of the results of that plant's PRA. Errors in that comparison-(such as. omission of the fact that Oyster-Creek has two isolation condensers vs. one at Millstone) affect its.results. Considering.the early state of the art of PRA at the time,the BNE work was performed (especially regarding containment failure and releases), and the errors and large uncertainties.in such a process of extrapolation from one plant to another,:no significance can be placed.on. the stated results. BNE CONCERNt "Of the 87 issues, some judgement of potential'rlsk-significance could be made on only 48..The remainder were related to external events not-considered in PRAs. Only 10 of the 48 were'found to have potential risk significance. Of the 10, backfit requirements are being imposed.on 9. A summary of these results as tabulated in the report are shown in' Table 1." GPUN RESPONSE: As stated, PRAs at the-time of the IPSAR and the BNE contractor review generally did not include external events,.thus PRA was not used to judge the risk significance of such issues at'that time. PRA approaches were used later to evaluate some external events, specifically tornado missiles and wind speeds. The BNE contractor review generally confirms that. issues of potential risk significance were identified for possible backfit t requirements. The contractor identified one issue of potential risk significance which was not subject to a backfit requirement. The IPSAR analysis (Appendix D page 25-27) concluded that the issue would'nct affect plant risk, and we concur with that judgement (as did the BNE contractor in t ( another part'of his work, Final Report Phase I.'page 6). e i C320467 '
.~ II. BNE In-house Review 'III.BNE SEP Tonic Comments The Department of Environmental Protection, Radiation Protection Programs (RPP), Bureau of Nuclear Engineering (BNE) staff has reviewed the Draft l SER, issued in October, 1989 by the NRC, for the OCNGS~FTOL. On the basis of the Draft SER (NUREG-1382) the NRC staff has concluded that the OCNGS can continue to be operated without endangering the health and safety of-the public. The BNE, however, racommended in their letter of December 11, 1989, that the SER, which is largely dependent upon the BEP, has several' areas where more study of the technical basis is needed. Specifically, BNE provided their concerns on the following five SEP topics and one non-SEP issue. 1) SEP Topic II-1.C " Potential Hazards or Changes in Potential Hazards Due to Transportation, Institutional and Military Facilities" 2) SEP Topic II-1.B: " Population Distribution" 3) 3EP Topic III - 5.B " Pipe Break Outside Containment" L 4) SEP Topic III-6 "New Seismic Floor Response Spectra"- 5) SEP Topic V-6 " Reactor Vessel Integrity" 6) 10CFR Part 50.44 - Hydrogen Recombiner Issue I r We have reviewed BNE concerns on each topic listed above and our response is provided below. 1) SEP Tooic II-1.C " Potential Hazards or Chances in Potential Hazards Due to Transportation, Ynstitutional and Military Fnf lities." i t BNE CONCERN "Since 1982, the areas immediately north and south of Oyoter Creek ] have been significantly developed and Route 9 has become a y thoroughfare, i.e., transportation route for' commerce as well asifor. j public use"' J GPUN RESPONSE It is our opinion that Route 9 is not used as.the major thoroughfare for north / south commerce due to the following impediments: Posted speed limits range'from 25-45 mph many with traffic lights. Road conditions and restrictions prohibit rapid transit. Commercial vehicles are allowed on the N.J. Garden State Parkway, a major highway that runa parallel.to Route 9, south of Exit 105, approximately 30 miles north of OCNGS. No industrial centers in area. No industries in close proximity to the plant site that are expected to use or store large amounts of explosive or hazardous material. Typical transported items are end-user products and consumables (e.g., retail products, food stuffs). 1 -C320467 ' 1
1 Thirsfero, we belicv3'thtt th) trentportaticn of hnznrdnus m-tcricic on U.S..
- c Route 9 poses no significant hazard to the OCNGS.
2) SEP Toole II-1.Bt "Pooulation Distribution" ] l BNE CONCERN..fSUKKARIZED): 1. Discrepancies in population projections exist between "RPP Base-L Document" and NRC SER. This would be " intolerable" since the State of New Jersey is responsible for emergency preparednese around Oyster-creek. 2. Over 70% of the total population resides in 4'out of the 16 sectors within.10 miles around the OCNGS and population densities exceed.most, if not all, siting criteria guidance published by the NRC. { ' 3. Whether the'OCNGS site meets the Exclusion'and Low Population Zone criteria as defined in 10CFR100. i 4. More accurate population center analysis should be' conducted'in 1990. GPUN RESPONSE: 1. "RPP Base Document" which we believe was generated by a BNE. consultant is not attached to their letter (although the BNE letter states'that excerpts from this document are attached) and thus it is difficult for us to comment.. However, we were assured by the New Jersey Office of Emergency Management (NJOEM) in our recent communication that'the NJOEM is completely confident that a full. evacuation of the 10 mile. Emergency Planning Zone (EPZ) can be accomplished safely and'promptly. The planning responsibility for evacuating.the'EPZ residents l rests with-the NJOEM. It should be noted that NJOEM indicated to us in our discussion that they were not. aware of the "RPP Base Document". .i 2&3. With respect to the statement "...over 70% of. the total population - resides in 4 out of the 16 sectors within'10 miles." This is understandable when one considers that half of the 10imile EPZ is covered by water. The Reactor _ siting criteria 1 (10CFR Part 100) states l in their definition for " Low Population: Zone" (10CFR100.3(b)).that i "these guides do not specify a permissible population density'or total population within this zone because.the situation may vary from' case to I case". It is not clear just what population density criteria BNE, I believes is exceeded. It is our opinion that the OCNGS meets the definitions and criteria for Exclusion and Low Population' Zone as given s in 10CFR100. 4. In 1987 GPUN employed a consultant (DRA) to initiate an update to the-evacuation time estimate (ETE). The DRA~ report estimates that the ~; permanent population of the 10 mile EPZ to be 105,159-and a combined permanent and transient population to be 181,001. GPUNlis planning to update ETE data when the publication of the.1990 census data becomes available in approximately:a year. It should be noted that in 1981 a Jersey Central Power & Light report estimated that "By 2010,1the population within a ten mile radius [of the OCNGS) will triple". Based on the numbers provided, the growth rate seen thus far closely j parallels this prgjection. 'l C320467 1
3) SEP Topic III-5B'" Pine Break Outside Containment". The NJ BNE staff's comments on the' physical configuration of_ isolation condenser steam supply line isolation valves and GPUN's proposed use of leak-before-break (LBB) criteria to address SEP Topic III-5.B, " Pipe Break Outside Containment", do not take into consideration the planned replacement of all applicable isolation condenser outside containment _ pipe with Type 316 Nuclear Grade Material. In addition, this project,_ scheduled for the next refueling outage, will also include replacenent of all four isolation condenser penetrations and the six outside containment isolation valves. The net result of this project will be a significant improvement in the quality and-safety of the plant.. A LBB evaluation ~will be performed on the replacement piping which will address all the latest criteria' developed by.the NRC Staff and given in draft Standard Review Plan Section 3.6.3. It will address the areas of concern expressed in the BNE study such as water hammer, fatigue and corrosion. GPUN believes that the probability of pipe rupture can be demonstrated to be extremely low so that LBB applies. The concept of LBB as a resolution to this issue has been proposed since 1980. At that time, and also recently, GPUN_ studied the j feasibility of installing an isolation valve inside containment on the f isolation condenser steam lines. Both studies concluded that l exposures, available space, accessibility, structural changes to the biological shield wall and costs preclude this modification as a-viable alternative. Thus, while the criteria for LBB are now more j stringent than during-the SEP review, it is an appropriate alternative in conjunction with the pipe replacement. - i q 1 The BNE identified specific-conclusions from their studies with-regard to the isolation condenser steam line valves. The conclusions and GPUN's response to them are as follows:- i i BNE CONCERNt "The present configuration of the isolation valves of the isolation. condenser line, does not meet the NRC General Design Criteria (GDC) on " Environmental and missile design basis" and on " Reactor coolant-pressure boundary penetrating containment". Neither a leak nor a j break could be tolerated at certain critical locations without 1 significant radiation release to the environment." GPUN RESPONSE: i The GDC were promulgated after Oyster Creek construction was completed. ['I and are not required to be met at OCNGS.- Meeting their intent is acceptable on a reasonable, defined basis.. Application of LBB can be that basis. On page 7 of the BNE report entitled "Off-Site Dose Consequences Resulting from a Pipe Break between the Drywell and Isolation Condensers at the Oyster Creek Nuclear Generating Station" (an attachment to the DEP letter of December.11, 1989) it is stated 1 that "When potential radiological accident consequences are analyzed using Nuclear Regulatory Commission (NRC) guidelines-for a single .{ C320467 failure accident, the resulting off-site doses are well within 10 CFR! l 100 criteria."'sThe BNE's own results suggest that'the current valve configuration is adequate for licensing consideration from a dose impact alone. Postulating. multiple failures =and severe accident source terms, as this study goes on to discuss is inappropriate for backfit or licensing decisions since these scenarios are outside of the scope of current NRC regulations. BNE CONCERN "The austenitic stainless steel exhibits leak-before-break under normal loading. However, under adverse loading conditions,' comprising of_ seismic and/or water hammer loads, the isolation condenser piping may have a major rupture." GPUN RESPONSE: As part of a. state-of-the-art LBB evaluation, GPUN will demonstrate-that the probability of a pipe rupture due to the loading conditions-identified above is extremely low.- BNE CONCERN 4 " Jet impingement of steam at operating temperature onto the valve j operator and its control _ circuitry, raises environmental qualification concerns for the isolation valves." GPUN RESPONSE: This issue is more fully described in the attachment to the BNE report as applying to steam impingement from leakage. Isolation condenser isolation valves are environmentally qualified for certain conditions as required by NRC regulations. This does not include consideration of direct steam impingement due to postulated leaks. The probability-of a steam leak of sufficient size and orientation to affect both isolation valves is very low. i BNE CONCERN l "A loss of coolant accident (LOCA) outside the drywell has several radiological consequences to the public, see report in the Attachment Section. An automatic isolation valve inboard to the containment would preclude a LOCA outside the primary containment." GPUN RESPONSE: The severe radiological consequences mentioned above are due to the BNE's postulated beyond design basis event. Consideration of this is j inappropriate for licensing. Quality pipe with extremely low rupture probability also precludes a LOCA outside containment. With respect to the off-site dose consequence evaluation performed by BNE (the attached BNE report) we found the following inconsistencies. C320467 i a
.I a. Item be on page 3.says that the fractional release remaining in the plume is from Reg. Guide 1.111. This probably refers to-a Figure 2 on page 17 of Reg. Guide 1.111.' The figure represents-the fraction of particulates and radiciodines remaining in a plume after dry deposition is calculated.- Item h on page 4 of the BNE report contradicts the-use of this figure stating that the only removal mode for radionuclides is radioactive decay, b. Release durations of eight hours for the. clad damage and ten. I hours for the fuel melt scenarios are stated on page 3. Page 4 then gives exposure = durations of two hours for.25 miles, eight hours at two miles,,and'30 dayo at five and ten miles. -Based on: the BNE assumptions for wind speed and meteorological conditions, release duration and exposure time should be approximately the same unless the exposed individual is moving with the plume. c. The calculations assume a constant' release rate.- This is arr invalid assumption because the release rate is a function of. reactor pressure. In fact, maintaining a significant release rate and damaging the fuel may not be possible..If'the break is small-enough to allow the reactor to stay at pressure, the leak rate would be small and reactor level would probably not be lowered enough to damage the fuel. Conversely, if the break is large enough to lower level sufficiently to damage fuel, in most cases reactor pressure would also decrease, eliminating the driving force for the release. Further, BNE calculation is based on exuremely high iodine to noble j gas ratio and multiple = failures of safety systems. In spite of the above, the BNE calculations show that the off-site doses are well-within the Federal standards. e SER_ CONCERN: i "The necessary restraints to prevent pipe whip, which is= required per NRC design criteria are not provided." U GPUN RESPONSEt 1 An acceptable LBB evaluation as required by the recent revision of GDC 4 will preclude the need to account for the dynamic-effects of a pipe . ) rupture. GPUN believes such an evaluation can be. performed and-found. acceptable. Therefore, pipe whip restraints are not' required. l 4) SEP'Toolc III-6 "New Seismic Floor Resoonse Soectra" BNE CONCERN "BNE staff wants to know what seismic criteria were used when OCNGS was designed and constructed, compared to the seismic criteria used in the design of modern nuclear plants in the eastern U.S." C320467,.,.. . -. I
q GPUN RESPONSEt When OCNGS was designed and constructed, seismic requirements similar to the design and construction of modern nuclear plants in the-eastern U.S. were utilized. The difference lies in the methodology and procedures that were used in seismic analyses..Obviously, modern ~ t nuclear plants have taken advantage of modern or " state-of-the-art" j concepts that evolved during the last' twenty years, At the time when OCNGS was designed'and constructed, procedures for seismic analysis were unsophisticated,-therefore, a very;condervative safety margin was included. For example,-piping evaluation utilized i 1/24 critical damping and the stress limits required by the codes were; well below the yield strength of materials. In contrast, modern power. i plants were designed based on a 3D dynamic model, realistic dampingg i l coefficients, and allowable stresses which are near or higher than.the: a yield strength of the material. l ? As more technically defendable design methodologies evolved, GPUN applied these to the analysis and evaluation of OC structures and components, such as the development of new floor seismic response spectra, 3D dynamic model and use of Code Case N-411 as was mentioned ~ in GPUN's November 1, 1989 letter to the NRC for the evaluation of t existing safety related piping and supports to satisfy the requirements of IEB 79-02/14. However, development of these new seismic response spectra require'NRC's review and approval which is an i l ongoing effort. It should be mentioned.that existing piping.and supports have been evaluated and modifications have been made utilizing new floor response spectra (developed in 1987 by Blume ^ Associates, GPUN consultant) that were developed from OCNGS SEP ground response spectra (generated by USNRC), satisfying the requirements of IEB 79-02/14. An independent seismic evaluation of OCNGS and-other nuclear plants in the eastern and central U.S. was done by the seismicity owner's Group (SOG) and EPRI. The study evaluated seismic hazards at:57 sites with i the objective of resolving the charleston Earthquake Issue. 'The study I has been documented in EPRI NP-6395-D, April 1989..The USNRC and its l advisor, USGS has reviewed this document and issued a Safety. Evaluation Report (SER) on September 20, 1988.
- i The study showed that OCNGS seismic hazard prior to the' Charleston:
Earthquake Issue is about the same or even better than other plants.as shown in Attachment 1 to this letter (Solid point, Site No.'43),-where the annual probability of exceeding the plant design acceleration.is approximately 2.0 x 10-5, to this letter depicts the seismic hazard of Oc exceeding 0.65 level, which is representative value of a typical plant median' acceleration - resistance-against core damage and prior to the Charleston Earthquake Issue..It is shown that for OCNGS.(Site No. 45), the annual probability of exceeding the acceleration level is approximately 2.5 x 10-6 which is about the same as the other nuclear plants. C320467 :
BNE CONCERN "The BNE staff recommends that the repair of the remaining 23 supports in question should not be postponed." GPUR.RESPONSEt Per GPUN letter of November 1, 1989, there is a high degree of j assurance that the existing condition of OCNGS piping and supports -(i.e., without the 23 modifications) has the ability to withstand a design basis seismic event without endangering their function.- The conclusion was made after the piping and supports were analyzed for requirements of IES 79-02/14 utilizing the new floor seismic response spectra ('87 Blume Spectra) that were developed from OCNGS SEP ground response spectre;(generated by USNRC), and " state-of-the-art" methods for calculating stresses in the components.- BNE's understanding of GPUN's November 1,.1989-letter to the NRC is that the modification of the, supports is difficult to implement-by 13R because of " schedule problems." The " schedule problem" stated here is for the NRC review and approval of.new floor seismic response spectra that is being developed by GPUN-and their consultants that will replace the " original FSAR criteria." l The new floor seismic response spectra being developed is a ] technically-supportable, " state-of-the-art" methodology that will be .j utilized to seismically evaluate oc structures, systems and components i including the remaining safety related piping and supporta per requirements of IEB 79-02/14. i 5) SEP Tooic V-6 " Reactor Vessel Intecrity" BNE CONCERNt "How does one know the most limiting material in the belt line region of the vessel and determine its reference Nul Ductility transition-i temperature to establish the Heat-up and Cool-down curves? GPUN RESPONSEt l GPUN has collected a significant amount of-material information from ) various sources in order to establish the initial unirradiated + properties of the beltline materials of the Oyster Cre'k reactor e vessel. In fact the chemical composition and impact properties for each beltline plate material is now known. Available materials information is compiled in Attachment 3 to this letter (TDR-725 Rev. 0). 1 The selection of the limiting material is also explained in Attachment 3. Basically GPUN performed an evaluation using both Regulatory Guide i 1.99 Rev. 1 and Draft Rev. 2. The 32 Effective Full Power Year (EFPY) fast neutron fluence at the inside surface of the reactor vessel was used in these calculations along with known chemistries for each plate material. This evaluation showed plate Number G-8-6 to be.the limiting material. C320467 b
'Tho initici rOforanco nil ductility;tamperaturo for occh'pleto material was estimated'from actual charpy-impact data. ' Attachment 3 l contains the impact data and initial reference. temperature values. With respect to the oyster. Creek 15EFPY P-T curves,-the1 adjusted reference temperature (ART) for;the~ limiting plate. material was extrapolated from actual measured material property values from the surveillance specimens removed during'10R..This' plant-specific ART was compared to that predicted by Draft Rev. 2 and was determined to be conservative with respect to the Draft Rev. 2 prediction. Thus a plant specific ART was used in the preparation of the P-T and hydro limits for oyster Creek. Attachment 3 provides the results of testing and analyzing irradiated specimens from the Oyster. Creek surveillance capsule. This report also describes in detail the calculation of ART for the limiting material. Several months after the oyster Creek 15 EFPY.P-T curves were approved by the NRC, Generic Letter 88-11 was issued and requested each licensee to perform'an evaluation of P-T limits using the calculational methods of the final Rev. 2 of Regulatory Guide-l.99. GPUN performed this evaluation and concluded that the limiting -j material in the beltline' region of ths' Oyster Creek reactor vessel is-l now an axial weld instead of a plate. The impact on the Oyster Creek P-T curves is a 4*F increase at 15 EFPY. The current Oyster Creek P-T-curves were developed using an ART of 125'F; therefore,'GPUN plans to prepare new P-T curves-in the near future. BNE CONCERN: j i " Provide the necessary information to perform an independent analysis of the effect of irradiation on OCNGS beltline materials" GPUN RESPONSE We assume that the BNE wants to perform a Regulatory Guide 1.99 Rev. 2 l calculation. to this letter.(excerpted from TDR-725 Rev. j
- 1) provides. sufficient information to perform such a' calculation for-l each beltline plate and weld.
{ l BNE CONCERN: 1 "The welds in the Reactor Vessel belt line region and other critical' locations such as nozzles have not'been inspected and can not be inspected per A.S.M.E. Section XI code C2 ring the life of the plant due to lack of accessibility. We understand that there was a waiver granted to the licensee on this inspection. Question: How does the vessel integrity get verified without the inspection of these critical areas? Please provide the BNE with additional information which led to the waiver being granted to the licensee." I 4 I C320467 _ _.......
t GPUN RESPONSE: 3 Relief (Waiver)-requests were granted for the 2nd inservice inspection (ISI) interval which ends in October 1991. These waivers were granted t at a time when the technology did not exist that would provide-reliable data in an "ALARA" manner. Since then, however, the technical and regulatory environments have changed resulting in the activities listed below 1. .BWR utilities, EPRI, and inspection contractors have been working' to develop tooling that will provide access to the vessel interior for volumetric inspections from the ID.(as is not done for most PWR vessels). For Oyster Creek, we anticipate that the j vast majority of.tho' beltline weld seams will become'" accessible" for the code-required examination. 2. The ASME B&PV Code Section XI have been' amended to require 100% examination of accessible welds in the beltline region.. NRC has made it clear to BWR Owners that'" blanket" relief-(waivers)'to these requirements willinot'be granted-in-the. future. As explained in Item 1, we expect to be able to examine most of the-beltline weld seams in the next 10 year interval to October, 2001. 3. We plan to inspect Control Rod Drive Nozzle Number 9.in the 13R outage (January 1991)._ The examination will be performed using state-of-the-art ultrasonic techniques. 4. Two recirculation-to-safe end nozzle welds were examined in 12R. Four more will be examined in 13R and the remaining four will be examined in 14R. All 10. welds will be regularly examined in the future in accordance with ASME Code requirements. 7 5. Other nozzle-to reactor vessel welds.will be re-evaluated for OD and/or ID accessibility as part of GPUN ISI Program Update Project. We expect to submit-our updated program to the NRC by l April 1991. Any future relief requests will be submitted with the update along-with technical ~ justification for these requests.
- 6. provides.information that led to-the waiver being granted to the licensee on welds in the reactor vessel beltline region and other critical locations such as nozzles.
l' 6) Hydrocen Recombiner (10CFR Part $0.44 - Standards _for Combustible Gas-Control System In Licht-Water-Cooled Power Reactors 1 i BNE CONCERN: "The SER does not address the requirement of 10 CFR 50.44 mmendment, requiring hydrogen recombiners. GPUN believes, based on a study performed by the BWROG and documented in NEO-22155 entitled, " Generation and Mitigation of Hydrogen in Inerted Containment,"- hydrogen'recombiners for the containment are not needed. The BNE staff believes that GPUN may be misinterpreting'the amendment requirements. Please provide clarification to the BNE on the'NRC's interpretation of the intent of the requirements." C320467 a f =~ w --
.s GPUN RESPONSE The draft NRC SER does address 10CFR50.44 and specifically the hydrogen recombiner issue on Page 6-12 of the SER. 10CFR50.44(c)(3)(ii) applies to plants which re'ly upon a purge repressurization system as the primary means of controlling combustible gases following a LOCA. Generic Letter 89-04, May 8, 1984, set out the Staff's determination that,-.for purposes of compliance with 10CFR50.44(c)(3), Mark I BWR plants will be found to-l not rely upon pressurization and purge system as the primary means of hydrogen control if three technical criteria are met. 5 Issuance of the generic letter followed the Staff's review and acceptance of the BWR Owners Group studies documented in NEDO-22155,- " Generation and Mitigation of combustible Gas Mixtures in Inerted:BWR Mark I containments", June, 1982. -The studies showed that, for all' BWR plants with inerted Mark I containments, peak oxygen concentrations following a LOCA are maintained below the combustible gas limits without requiring venting or hydrogen recombiners. GPUN demonstrated compliance with the three technical criteria for Oyster Creek in its letters of July 13, 1984, August 14, 1985, and June 16,: 1985.' Relevant information was also provided in GPUN letters to the Staff dated August 2, 1982, December 15, 1982, and September 24, 1985.- GPU Nuclear does not believe they are misinterpreting the regulatory requirements as identified in the BNE comment...By letter dated November 15, 1988, the Boiling Water-Reactor Owner's Group submitted I their compliance position regarding 50.44 for several utilities with inerted Mark I containments. GPU Nuclear's position is consistent with that submittal. I
SUMMARY
i It appears that most of the BNE concerns expressed in their letter of December l 11,'1989 are based on some misunderstanding of the issues (e.g., seismic floor l response spectra, hydrogen recombiner, population distribution issues) and/or l lack of detailed knowledge of the GPUN's future plans (e.g., Isolation l Condenser piping replacement, reactor vessel integrity issues).. Our responses provided above should clarify some of their concerns. Our review of the BNE l commente did not raise any new concern that would refute.the conclusion reached by the NRC staff - that OCNGS can continue to operate'without endangering'the health and safety of the public. ~i Very truly yours, f <;u e-R. L. Long Vice President, Corporate Services RLL/cjg j t C320467 l
~ cc: Mr. William T. Russell, Administrator Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. Alexander W. Dromerick U.S. Nuclear Regulatory Commission Washington, DC 20555 NRC Resident Inspector Oyster Creek Nuclear Generating Station Mr. John F. Stolz, Director Project Directorate I-4 Division of Reactor' Projects -I/II office of Nucear Reactor Regulation US Nuclear Regulatory Commission Washington, DC 20555 Mr. Kent W. Tosch, Chief Bureau of Nuclear Engineering Division of Environmental Quality N.J. Department of Environmental-Protection CN 415 Trenton, N.J. 08625 1 I l l I i i i I 4 3 I i i l i i . i i C320467 3
~ . ~ tlt ATTACHMENT 1. (From EPRI NP-6395-D, April 1989) r l l A AND B HAZARDS AT PLANT SSE ' ACCELERATION I l Results-for 57 sites-(59 units)- L 10-a C3 O I 210-r c O 0 -o. . o ci3 g P " o U10-' /*>og '" .o .lWh> - 9 "l o o 5 i g oooo o o " 14) " "jp/ l -o *** o ,, g a ,m o o a5;o o e o oo o-o 10-5 n r oo ca o o o - o' o o oo .o a o o. -o $10-s o 7 2 i az 10-7 r l 2 i Analysis A, Median Hazard i 4 . o Analysis B, Median Hazard + 10-8 '. ' ' 3 0.' l 0. 10.' ' 20
- 40. ;
5 0.- 6 0. -. SITE NUMBER- .e oc sms Figure 6 2. Seismic he.zard results at plant SSE acceleration, for Analysis A and B. (Error _- bars denote the range of 15* and 85'h percentile hazards for Analysis B). 'f j i I i i i l-6-10 i 4 y i$
- A- -
r- -,
.q ATTACHMENT 2 (From EPRI NP-6395-D, April 1989) A AND B HAZARDS AT 0.65g Results for 57 Sites 10 g { g . -Analysis A Median Hazard o 10_4 o Analysis B Median Hazard-r 2:< -1 Q $10-5 r u .o. j, 1 ot[g y ." o o-w on o. "o gp ", 10-' o o 'lg g f*o i ,,i,oo" l g onoo o o ', m o g g ,o o o m o .o on o >o ci 10-7 r oo o o o o o o o o o cc o o )10-8 r 1 g g y $10-' r i 1 l 10-10 O. 10. 20. 30. 40. t 5 0.' 80. SITE NUMBER-C OC Srrs Figure 6 4. Seismic hasard results at an acceleration level (0.65 ) representative of a typical 9 plant median. acceleration resistance against core darnage, for Analyses A and B. (Error bars denote tfee range of 15* and 85* percentile hazards for' Analysis B). i p" 6-13 ? 'k
\\ &. ar ATTACHMENT 3 4 S TDR NO. 725 REVIS40N NO. O TECH.NICAL DATA REPORT $NvNY NO. BA328011 PAGE I OP 193-PROACh Oyster Creek Nuclear DEPARTMENT /SECTION E&D/NA&F Generating Station RELEASE DATE 4 '2(="M REVISION DATE DOCUMENT TITLE: Testin; and Evaluation of Irradiated Reactor Vessel Materials Surveillance Prozram Specimens CRIGINATOR SiqNATURE DATE APPROVAL (8) SIGNATURE 'DATE AfA]/ /O 22.M - Nd// ^ ^ ^ N /e/n1)sa -* N/Vm~ /b -22 M - y APPSVfL PgR EXTERNAL DISTRIBUTION DATE ANLJ wtd AT4539 l Does this TDR include recommendation (s)? GYes ONo if yes A/TR # e DISTRIBUTION ggt: l lyf 0 54 N Tne Oyster Creek Technical Specification requires that new Pressure-Temperature (P-T) operating curves be developed for operation beyond 10 Effective Full Power Years (EFPY).- This report providee.ne results of testing and analyzing-irradiated spec vis removea from RVMSP Capsule No. :2. The results are uses :;o predict the reference nil-ductility temperature (RTn97) for future periods of operation. The l cajusted RTNDT will be used for developing new P-T curves. ...d. Tests / Analyses' Performed Soeolmans representing the plate, weld,- and heat affectec zone (HAZ) were subjected to Charpy V-notch" impact tests, tensile tests, and chemical analyses. Dosimetry' wires were analyzed to determine neutron flux and fluence. Tne radjusted RT,3pi values were precicted following the general -recommendations of Reg. Guide 1.99 Draft Rev.' 2. Key Results Tne fluence 'at the capsule was 7.46x10I7n/cm2(E >1MeS). The measurea shift in the Charpy test 30 ft-lb. temperature was.720F for the plate material which'was,limitin . measured shif t exceeded that predicted by Reg. G;g.. The., 1de l.99, l,Rev.1 (400F) and Draf t Rev. 2 (580F). There was i essentially no reduction in the Charpy upper shelf energy l The adjusted RTN for various periods of operation was:.. 8.38 EFPY-96 0F,DT10 EFPY-1040F,15 EFPY-1250F, 20 EFPY-1410F, 24 EFPY (End-of-License)-1540F, 32 EFPY-1700F. The predictec EOL shift in RTNDT is less than the maximum permitted in 10CFR50 Appendix G. Future Actions 1. Develop new P-T curves and submit to NRC by 2/28/85. 2. Reinstall surveillance capsule in cycle 12R outage and remove and analyze at approximately 15 EFPY. { oCOVER PAGE ONLY gg 'g-Accoooso osa
TDR'725' Rey, 0 - 4 i.- Page la t TABLE OF. CONTENTS- - PAGE
1.0 INTRODUCTION
2 2.0. VESSEL BELTLINE MATERIAL DESCRIPTION 3 3.0 ~ SURVEILLANCE PROGRAM DESCRIPTION 7 4.0 TESTLMETHODS .10' 5.0 TEST RESULTS 12 6.0 DISCUSSION 15 7.0 RT AND CV PREDICTIONS 18-NDT USE j
8.0 CONCLUSION
S 26 9.0 FUTURE ACTIONS 27'
10.0 REFERENCES
'29~ i 11.0 TABLES 30. l 12.0 FIGURES 40 13.0 APPENDICES 46 TOTAL EFFECTIVE PAGES 193 1605H ? .N-e ---n.-- ~ s
'i 4 ,1 TDR 725 Rev. 0 j Page 2 11 -]
1.0 INTRODUCTION
1 1.1 Purpose j. The Oyster Creek Technical Specification, Section 3.3.A (iv), re-I l quires new pressure - temperature (P-T) limits for operating the reactor.beyond ten effective full-power years (EFPY).. This report provides the results-of evaluating the test data generated'from materials contained in the Oyster Creek Reactor Vessel Materials Surveillance Program (RVMSP) Capsule No. 2. The' test results form' the basis for generating the new curves. 9 1.2 Backaround 1.2.1 General i Ferritic reactor vessel materials that surround the active core experience a reduction in fracture toughness when exposed to neutron irradiation. The extent of reduction depends primarily upon material l-chemistry and fluence. In order to assure adequate margin against l brittle fracture of the reactor vessel, fracture toughness require-ments are specified in 10CTR50 Appendix G[1}. Requirements for' monitoring the effects of-radiation on specific vessels are specified 5 in 10CFR50 Appendix H{2) which requires periodic testing of reactor i i a vessel material samples removed from the reactor after exposure to i actual operating conditions. ( 1 1605H I
TDR 725L 'o ( ' Rev. 0 Page 3 t The NRC has also issued Regulatory Guide (R.G.) 1.99[3] which pro-vides guidance to determine the reduction in toughness properties as a function of fluence and asterial chemistry-(copper, phosphorous, nickel). The toughness properties addressed in-R.G.:1.99 are
- 1) shift in NDTT and 2) reduction in upper shelf energy..
1.2.2 0Fster Creek - L In order to seet-the requirements of'the: Technical-Specification and 10cFR50 Appendix H. GPUN removed RVMSP Capsule No.'2 following 8.38 s i EFFY. This capsule contained material test specimens'(manufactured:' from materials used to fabricate the-vessel)-and dosimetry' wires. The specimens and wires were tested:and analysed; the results were-I j. evaluated to provide the basis for generating new P-T curves. { ~ 2.0 VESSEL BELTLINE MATERIAL DESCRIPTION 2.1 Introduction This section provides all the available data and information for the Oyster Creek vessel beltline materials. It is noted that since the Oyster Creek vessel was fabricated before the issuance of Appendices G and H of 10CFR50, much of the currently-required data was not gen-erated. This subject was discussed in detail in NUREG-0569[4]. S 1605H l
t-i TDR 725 i Rev.-0 Page 4 2.2 General Information The vessel was fabricated by Combustion Cr.gineering in accordance with General Electric specification 21A1105.- This specification 4 included requirements for fabricating test specimens for use in a j surveillance test program. GE fabricated'and assembled three sur-. j veillance capsules which were installed on-site. 2.3 Beltline Material Location The location of each beltline plate and weld is shown in Figure 1. Material heat and lot numbers are provided in Table'1. 2.4 Beltline Material Description 2.4.1 Plate Plate was ordered from Lukens Steel to the requirements of ASTM'A302 Grade B modified with the addition of 0.40-0.70% nickel.(Grade C in-current versions of A302). Combustion Engineering heat treated each' plate (6), after hot rolling, at 1550*-1600'F for-4 hours followed by water quenching and a 1225 25'T temper for 4 hours. 1 2.4.2 Weld Material f-Submerged arc welding of the six longitudinal welds and one circum-ferential weld was performed with Raco #3 bare wire and Arcos B-5 flux. Manual metal arc welding was performed with 8018 covered electrodes. 1605H t --- = -.
4 'TDR 725 Rev. 0 Page 5' 6 2.5 Vessel Fabrication Post-Weld Heat Treatment (PWHT) l Interstage PWHT's of 1125-1150'F for 15 minutes, minimum, were per-formed; the number of cycles is unknown.- l l l The final vessel PWHT was performed at 1150 25'F for 10-1/2 hours, 1 i l 2.6 Toushness Requirements l l Tha beltline materials were required to have nil-ductility transition temperature (NDTT) of +10'F, or less, as verified,by obtaining a Charpy V-notch absorbed energy of at least 30 f t-lbs at +10*F. i Additionally, Charpy V-notch transition curves were required for each-l beltline plate. E l e 2.7 Beltline Material Data 2.7.1 Test Specimen Location and Orientation 2.7.1.1 Plate Tensile and Charpy specimens were removed from the 1/4T (T= thickness) location in the plate. The long axis of the specimens were parallel to the plate rolling direction. The Charpy specimen V-notch' was [ ' machined perpendicular to the plate surface. 2.7.1.2 Weld The Charpy specimen long axis was perpendicular to the weld direction-and parallel to the plate surface..The middle of the specimen con-tained the weld and the notch was centered in the weld parallel?to the weld surface. 1605H-i 4
1 TDR 725 'Rev. 0 Page 6 The tensile specimen long axis was parallel to the weld direction with the gauge length containing only weld estal. i q 2.7.2 Plate Properties q The available data for each plate are provided in Appendix A.L. The chemical analysis results are-as reported by Lukens Steel.- Combustion Engineering developed the tensile and Charpy data using-material removed from each plate following heat treatment and a sie-ulated PWHT of 1150'F for 30 hours, j The maximum NDTT among the six plates, as determined.from the tran - Lj sition curves, was +10'F. j I Copper was not reported as it was-not required for acceptance.- l Phosphorous contents ranged between 0.006% and 0.019% as reported by- ~ Lukens Steel. As can be seen. Lukens Steel did not report nickel for heat number T-1937 (two plates). One of the plates f rom this ' heat was used ~ in the surveillance program. Chemical analysis of surveillance speci- ~ mens identified a nickel content of 0.107%. The reasons for these plates' being accepted and used in the vessel are unknown. 1
- }
The discrepancy has no effect on the design and operation of the reactor vessel. These two plates met all the-mechanical and Charpy-test requirements of the initial material specification. The ASPE B&PV Code allowable stress values are identical for ASIM'A302 Grades B and C. ~i 1605H
l + TDR 725 y 1Rev. 0 Page 7 l 2.7.3 Weld Material The available data for the submerged are welding materials are. pro-vided in Appendix A.2. Combustion Engineering developed the data i i from specimens removed from test welds subjected,to a_ simulated PWHT i-of 1150'125'T for 40 hours. ll Charpy data are reported for three of the four combinations of wire a and flux. The minimum average value at +10' was 57 f t-lbs.. Tran-sition curves were not developed for weld materials.. t i The copper content of the weld deposit was not determined since it i was-not required for acceptance. The copper content of one. heat of i bare wire was reported; however,'since most low alloy steel wire was copper coated for corrosion prevention, this data is essentially meaningless with respect to the copper content of the weld deposit in i I that the copper coating becomes a part of the deposited weld metal.- [ The phosphorous content ranged-between 0.013% and 0.018%. c There is no data available for the 8018. electrodes. 3.0 SURVEILLANCE PROGRAM DESCRIPTION i. 3.1 Introduction l. The Oyster Creek surveillance program was evaluated by the NRC as part of the Systematic Evaluation Program (SEP) and found to be l acceptable (NUREG-0569). Because the Oyster Creek program was i l initiated before the issuance of 10CFR50 Appendix H, the program does not meet all the requirements of the Appendix. 1605H l
TDR'725: l t ~Rev. 0-Page 8 l 3.2 Surveillance Capsules Three surveillances-capsules were installed in'the reactor vessel at ~ the 30', 210', and 300' locations. The specimen holders were located at the mid-plane of the active fuel (Figure 1). The type and number of specimens in each capsule is listed in Table: 2. I The 30' capsule was removed in September 1971 due to an installation j error that made removal of the capsule necessary to retrieve'the1 4 dosimetry wires. Its specimens are currently in storage. j The 210' capsule (RVMSP Capsule No. 2) was~ removed in March 1984 3 after 8.38 EFPY. The specimens from this capsule were tested and analyzed by Battelle Columbus Laboratories to determine the effects i of radiation on the Oyster Creek beltline materials. 9 3.3 Surveillance Material Description j The surveillance material test specimens were machined from plate i l heat number T-1937, slab 1 (referred to as T-1937-1), wire heat num-- .t ber W5214, and flux lot number SG13F. l t The plate was part of a recirculation outlet nossle cutout from piece number 564-03E (Figure 1). This was also one of the plates exhi-biting an NDTT of +10'F as determined from a transition curve (Appendix A.1). i t c 1605H .,_.= +
i-TDR 725 Rev.~0 Page 9-The weld materials used in the surveillance program were not used in the beltline welds; however, they were manufactured by the same sup-pliers as those used in the beltline (i.e., RACO #3 wire,'Arcos-B5- ~ flux). Deposited weld properties are not expected to show a large variance as evidenced'by the test results shown in Appendix A.2. Following welding, the assembly was subjected to a simulated stress relief of Il50 25'F for 30 hours. 1 3.4 Test Specimen Description 3.4.1 General Charpy specimens were standard-sized meeting the requirements of ASTM i { E23[5]. The tensile-specimens ~were 1/4-inch diameter, 1-inch gauge length, meeting the requirements of ASTM'E8[6]. I 3.4.2 Plate and Weld Specimens t i The plate and weld specimen location and' orientation were the same as I described in 2.7.1.1 and 2.7.1.2, respectively, q i 3.4.3 Heat Affected Zone (RAZ) Specimens j i The Charpy specimen long axis was perpendicular to the weld direction i and parallel to the surface. The notch was parallel to the weld. direction with the base located at the edge of the weld. i e 1605H
l 4 TDR 725 + Rev. 0 Page 10' The tensile specimen long axis:was perpendicular to the weld direc-t tion and parallel to the plate surface. The RAZ at one edge of the ~ weld was centered in the specimen gauge length. 3.5 Caosule No. 2 Contents This capsule contained 10 tensile' specimens, 36 Charpy specimens,1and 3-esch copper, iron, and nickel dosimetry wires (see Table. 2); Two of the tensile and nine of the Charpy specimens were identified i as " APED Standards." GE could not determine the identity of these. specimens. Since there was no baseline data, these specimens were not tested; they are currently in storage. l I i ' i j 4.0 TEST METHODS l 4.1 Introduction This section summarises the test. methods used.by Battelle Columbus Laboratories. Details of the methods are provided in Appendix B. 4.2 Flux and Fluence Determinations t The dosimeter wires were removed from the Charpy packets and cleaned until the dark oxide layers were removed. The wires were.then gamma-ray counted and radionuclide activity data were generated. Gamma radiation from.the dosimetry wire was measured and used to calculate the flux required to produce the level of-activity. The { fluence was then calculated from the. integrated power output of the o reactor during the exposure interval. 1605H
.~ TDR 725 L 1. Rev.~0 l Page 11 From the generated data, neutron flux and fluence values with-energies greater than 0.1 MeV and' greater than 1'.0 MeV at the i capsule, vessel' inner wall, AT,'and IT locations'were calculated. Displacement per atom per second (dpa/sec) values were also generated to have the information available should dpa becoes a factor in damage predictions.. The asinuthal location whichireceives-the highest flux was determined. t 4.3 Chemical Analysis Two samples of each of the plate and weld materials were-analysed for-copper, nickel, and phosphorous. Each sample was. removed from a. j broken half of a separate Charpy specimen. Wet chemical analysis'was performed by inductively coupled' plasma-optical emission spectrometry. One sample in each set was re-analysed. Standard solutions of appropriate concentrations were used'for calibration. 4.4 Charpy Testing i Charpy V-notch tasting was performed in accordance with ASTM E23' [5] using an instrumented impact machine. The data was-.used to generate plots of absorbed energy vs. temperature, percent shear.vs. tem-- perature, and mils lateral expansion (MLE) vs. temperature. :The 30 l and 50 ft-lb and 35 MLE temperatures and the upper shelf energy were established for each of the plate, weld, and RAZ materials. 4.5 Tensile Testing Elevated-temperature tensile testing was performed in accordance with ASTM E21[7]. One each of the plate.and weld specimens were tested at
- 1) the temperature of the onset of the upper shelf, 2) 550'F, and 1605H o,.
i TDR-725 '.r R::v. O i Page 12 l J .l
- 3) a mid-transition region. lOne each of'the IAZ specimens were j
tested at' the temperature of the onset of' the upper shelf and 550'F. 1 i Load-elongation curves were automatically recorded for each test. Data generated from the tests. included 1) yield strength; 2)= ultimate I strength;.3) fracture load, strength,'and stress; 4) uniform elon-- l-sation; 5) total elongation; and 6) reduction of area. 5.0 TEST RESULTE 5.1. Introduction This section summarises the key results of the testing described ~in 'Section 4. Detailed results are provided in Appendix B. l I i I It is noted that Batte11e's calculations for EFPY are based on a full Power valve of 1930 MWTh; GPUN's method of-calculating EFPY are based upon the license-allowed maximum powers during early operation.. As a result, GPUN's calculations result in-a higher va'lue of EFPY (8.38 i vs. 8.15). This difference has a negligible"effect on fluence and damage predictions. 5.2 Flux and Fluence Determinations i l The capsule flux and fluenceivalues for E) 0.1 and E > 1.0 MeV determined from the capsule dosimetry wires are provided in Table 3. The average full power flux and' fluence are listed below. Full Power Flux-Fluence Energy (n/cm*/see x 10') (n/cm* x 10 " ) 1 j > 0.1 MeV 5.15 1.32 ) 1.0 MeV 2.90 .746 a 1605H i . ~.
.1 l TDR' 725 Rev. 0 Page 13 i 1 Calculations revealed that the peak flux occura ~at the' 223.5' ) i asinuthal location. The flux distribution near the core midplane ~was fairly flat and the axial factor used is 1.017. The axial fact'r was u 1 applied to all the peak wall flux and: fluence results reported. j l 3 i i The flux and fluence values determined for E > 0.1 and E > 1.0 MeV at I i. t j the 223.5' location are provided in Table 4. Values for the' vessel surface, 4T, and IT locations are shown. As can be seen, i the vessel.inside wall is " leading'" the capsuli but the iT and i IT locations are " trailing" the capsule'in terms of exposure; the respective lead factors (i.e., the. ratio of flux (E > 1.0 MeV) ats i the capsule to the flux at the vessel wall location) are 0.76. 1.23; t and 4.62. For the purpose of estimating shifts;in RTwor at the AT location, this result indicates that the shift in-the capsule: specimens will be greater than the shift at.the AT location. ~ l 5.3 Chemical Analysis-The results of analysis of the surveillance materials.are'shown below. The results of the Lukens Steel (plate) and.C-E (weld) l analyses.are listed for comparison. j Composition (WT %)' Cateaory Cu P Ni Plate (Battelle) 0.17 0.011 0.11 Plate (Lukens) . NR: 0.011 NR i Weld (Battelle) 0.28 0.022 0.05 Weld (C-E) NR 0.018 'NR NR = Not reported / analyzed l t 1605H w w w -.-g9-g ,a%,,u ____,,,__________.____,__,m____________a_____._
I i TDR-725 l Rev. 0 Page 14 l i 5.4 ghgrey Test i . The 30 f t-lb, 50 f t-lb, and 35 M12 temperatures and upper shelf r energy for each category of test specimens are shown below. These f 1 data were obtained f rom transition curves developed as described in l ./ [ 4.3. The available data from the unitradiated plate material is 1 shown for comparison. Individual specimen results are provided in Tables 5, 6, and 7. Impact energy transition curve.s are shown in Figures 2, 3, and 4 l t Temocrature (*F) Upper Shelf ( Catemory Condition (*]~ 30 ft-lb 50 ft-lb 35 MLE Eneray (ft-lb) Plate U 15.5 53.5 N.G. 86.0 Plate I 87.5 150.5 97.0 85.5 Weld I 27.5 93.5 50.5 83.0 i ( HAZ I 32.5 64.5 43.0 74.5 3 l l .3
- - U = unitradiated I = irradiated NG = not generated e
t 5.5 Tension test The tension test results are provided in Table 8.. The unirradiated I base and weld metal results are listed for comparison. In all eight I specimens, necking occurred in the gage length. All failures were in a ductile cup-and-cone mode. 1605H -,a ..z. e _... _,,....--,-.,.-
1 1 4 TDR 725 i Rev. 0 Page 15 i f 6.0 DISCUSSION 1 6.1 Introduction l This section provides a discussion of the surveillance test results including a comparison to the expected, or predicted, results. -} l 6.2 Flus and Fluence Determinations The reported values fall within 0.5% of the expected values based on 'I k the analysis performed [8] on the dosimetry wires removed from the vessel in 1971 at the end of Cycle 1A. Analysis of those wires l Indicated that the fast neutron fluence per megawatt day was 1.29 x l 10 nytt analysis of the Capsule No. 2 wires indicated a value of 1.31 x 10nyt. 6.3 Chemical Analysis 6.3.1 Plate The copper content of the surveillance plate material (0.17%) falls l within the range of 0.15-0.20% as expected based on GE discussions with Lukens Steel (9). The phosphorous content (0.0111) is the same as reported by Lukens. i 4 i As noted earlier, the nickel content of the surveillance plate ~ material (0.111) is not in compliance with the order requirements P (0.40 - 0.70%). This result is consistent with the fact that Lukens Steel did not report nickel content for this best of steel. Again. the reasons for Lukens' shipping and C-E's accepting this heat of plate are unknown. i e 6 1605H , - - r ,v -.--,-,,,-n e .+ . ~,
TDR 725 Rev. 0 Page 16 ) i l The addition of nickel will increase the hardenability and toughness i of the material. As shown on the transition curves (Appendix A.1), the toughness of this heat is consistent with that of the other four heats with nickel added. Also, the lack of nickel has no effect when l performing evaluations on toughness to R.G.1.99 Rev.1. I 6.3.2 Weld I The copper, phosphorous, and nickel contents of the surveillance weld deposited are consistent with both expected and reported results. l As reported in (10), the expected copper content would be approxi-4 l mately 0.27% for welds deposited with copper-coated wire. The j reported value of 0.281 compares well with the expected value. s 6.4 Charpy Tests 6.4.1 Limitina Catenory \\ a 2 valuation of the Charpy transition curves leads to the conclusion t 2 that the plate material will be limiting with respect to RTwot. After 8.38 EFPY, the 30 ft-Ib temperature of the plate exceeds that of the weld and KAZ by 60* and 55'F, respectively. It is not ex-pected that the RTwor of the weld and RAZ will anceed that of the plate through the remaining life of the vessel. The limiting category for upper shelf energy is the EAZ. 6.4.2 RTwot Temperature and Upper Shelf Enermy Shift 6.4.2.1 RTwot The discussion on the shifts due to irradiation will center on the plate material only for two reasons. 1605H
) TDR 725 Rev. 0 Page 17 i 1. The plate material is limiting with respect to RTwot (see J 6.4.1), and ) 4 ) i 2. The only unitradiated data (i.e., Charpy transition curves) available is from the plate material. The temperature shif t of the capsule specimens as measured at 30 i ft-lbs is +72'F. Using R.G.1.99 methods and the surveillance test program fluence and chemistry results, the predicted shift is 40'F. i j It is noted that R.G. 1.99 prediction methods are considered valid i only for calculated shifts greater than 50'F. 4 l ~ r 6.4.2.2 Upper Shelf Enerzy (CVost) The measured shift in CVose of the plate is -0.5 ft-lbs (-0.6%). Using R.G. 1.99 F.fgure 2 and extrapolating downward for fluer.ce, the predicted shift in CVost is -12 ft-lbs (-14%). 6.5 Tension Tests D Neutron damage of low alloy steels results in hardening (i.e., strength properties will increate, ductility properties will decrease). Comparison of the unirradiated and irradiated data P results in the conclusion that there is only a slight effect of~ irradiation on the tensile properties. 0 1605H- )
_~ l TDR 725 Rev. O i Page 18 ) i i 7.0 RTwor and CVver PREDICTIONS l 7.1 Introduction J To develop P-T curves for an operating period, the adjusted RTwor j at the end of that period must be conservatively and reasonably i estimated. This section describes the methods used to predict the 1 adjusted RTwot, and the results, for several operating periods including end-of-license (EOL). L { Also, the effects of irradiation on the Oyster Creek beltline. j materials' CVues are assessed. i 7.2 RTworPredictions 7.2.1 Limitina Material i Selection of the limiting material (i.e., the material which will .i exhibit the highest RTwor at the end of the operating period) depends on two factorat 1) initial unitradiated RTwor and 2) material chemistry. The adjustment to made to RTwor is defined in 10CFR50 Appendix G as "the temperature shift measured at the 30 ft-lb j (41J) level in the average Charpy curve for the irradiated material relative to that for the unirradiated material". Selection of the i limiting material will focus on the plate material, only, for two t i reasons:
- 1) there is no unirradiated RTwor data available for:the weld and KAZ materials and 2) the 30 ft-lb temperature of the-irradiated plate exceeds that of both the weld and EAZ by 60' and 1
55'F, respectively. It is not expected that the RTwor of the weld and HAZ will exceed that of the' plate by the time the next capsule is removed (approximately 15 EFPY). 1605H h 4 -, - +.. _.. -,_....._._-_.___. _ _ ~ - -- _ -__ _ ---._ - _ -___-
- j TDR 725 Rev. 0 ] Page 19 1 -The initial unirradiated RTwer and the essential chemistry content ) (Cu and P for R.G. 1.99 Rev. 1 and Cu and Ni for R.G. 1.99 Draft Rev. 2 evaluations) are sununarised belows i 1 i i Initial Code R_Twot Cu ' r Ni G-308-1* +15.5 0.173' O.011' C 107 ' Same Beat 8 8 G-307-1 +15 0.173 0.011' O.16- (T-1937) t C-307-5 -19 NR 0.019 0.53 G-8-6 +12 NR 0.013 0.51 G-8-7 -4 NR 0.019-0.48 G-8-8 +5 NR 0.006 0.46 {-
- Surveillance capsule plate.
' Surveillance capsule results. An evaluation of the ARTwer at a 32 EFPY fluence of 2.38 x 10 ' n/ca" (E > 1 MeV) was performed for each plate. The 8 evaluation was performed using both Rev. 1 and Draft Rev. 2 methods i using the known chemistries. Where the copper content was unknown, a value of 0.20% was assumed (see Section 6.3.1). The results are provided in Table 9. s The limiting plate is G-8-6 based on both the Rev.-1 and the Draft Rev. 2 evaluations. I . h 1605H -, + - w, em-es* -e~. w v = -,w--,, -*+-w sw n--. v---
~ -.. TDR 725 Rev. 0 Page 20 7.2.2 Predictive Methodolony 7.2.2.1 R.G. 1.99 Rev. 1 The equation for predicting the shift in RTwor is A RTwor = CF (f)8'8 (1) (Paragraph C.la of Rev. 1) j where CF(Chemistry Factor) = 40 + 1000(ICu-0.08) + 5000 (EP-0.004) i F(fluence factor) = fluence (E > iMev) at AT + 10 ' 8 ECu = weight percent coppert if less than 0.08, use 0.08 j EP = weight percent phosphorous; if less than 0.008, use 0.008 The logic to be used to predict ARTwor for plate G-8-6 is as i follows: 1 1. Calculate CF using equation (1) for the surveillance plate (G-306-1) for the measured shif t (72'F) and fluence (7.46 x 10 ' n/ce' (E > 1MeV)). 8 2. Increase CF calculated in 1 by the difference in the CF's for G 9-6 and G-308-1 from Table 9 (37'F). 3. Calculate aatwor for G-8-6 using CF from 2 and fluence (E > 1MeV) at IT for appropriate operating periods using equation (1). e 4 1605H o
I TDR 725 Rev. 0 Page-21 I 7.2 2.2 R.C. 1.99 Draft Rev. 2 The equation for predicting the shif t in RTwot is (CF) f'*8'*810' (2) A RTwot = } where 1 CF(Chemistry Factor) from Table 2 of Draf t Rev. 2 j F(fluence factor) = fluence (E > INeV) at AT + 10 l j This methodology differs slightly from that recommended in Draft Rev. 1 2 in the following manners: 1. Draf t Rev. 2 permits the use of plant-specific data only when two or more surveillance data points are available. Since the i measured shift for the one Oyster Creek data point exceeds that predicted using Draf t Rev. 2,,it is more technically alp 4)priate l to utilise that point rather than using Draf t Rev. 2 predictive methods alone, i 2. The calculated fluence at iT based on the surveillance program dosimetry and calculations was used instead of the Draf t Rev. 2 attenuation factor because it was considered to be more representative to use plant-specific rather than generic data. p 1605H }
) TDR 725 Rev. 0 Page 22 3. The margin ters (equal to the predicted shift or 34'F, whichever is less) recommended in Draft Rev. 2 to be added to the calculated ARTwer will not be included. Since the Oyster Creek data point falls outside the data base used to develop the 3 margin ters, its use in the Oyster Creek predictions is l .onsidered purely arbitrary with no technical basis. I i The logic to be used to calculate ARTwer for plate G-8-6 f.s as follows: 1 1. Calculate CF for plate G-308-1 for the measured shift (72'F) and j fluence (7.46 x 10"). J i j 2. Increase CF calculated in 1 by the difference in the CF's for G-8-6 and G-308-1 from Table 2 (59'F). i 3. Calculate ARTwer for G-8-6 using CF from 2 and fluence (E > 1MeV) for appropriate operating periods. 7.2.3 ARTwerPredictions 7.2.3.1 R.G. 1.99 Rev. 1 The ARTwot for plate G-8-6 using the methods of J.2.2.1-is added to the initial RTwot (+12'F) identified in 7.2.1 to obtain the adjusted RTwot (ARTwor). The results are tabulated below! 1605H ,.,,_.---w ,.O, _. _., +, -...n.,- .n.. ,,,,-.w,,,-
i TDR 785 Rev. 0 Page 23 Fluence (E > 1MeV) Operating at IT Period (EFPY) CF (n/cm* x 10 ') ARTwer + RTwot 1 ARTwot = 8.38 300 0.61/ 74 + (+12) 86 = 10 300 0.73' 81 + (+12) '93 = 15 300 1.11 100 + (+12) 112 = 20 300 1.48, 115 + (+12) 127 = 32 300 2.38 146 + (+12) 158 = 7.2.3.2 R.G. 1.99 Draft Rev. 2 The ARTuor for G-8-6 calculated using the methods identified in \\ 7.2.2.2 is added to the initial RTwer identified in 7.2.1 to obtain the adjusted RTwot (ARTwor). The results are tabulated below: l Fluence (E > 1MeV) Operating at IT j Period (EFPY) CF (n/cm* x 10) ARTwor + RTwor ARTwer = 8.38 259 0.61 84 + (+12) 96 i = 10 259 0.73 92 + (+12) 104 = 15 259 1.11 113 + (+12) .= 125 20 259 1.48 129 + (+12) t = 141 32 259 2.38 158 + (+12) 170 = The current P-T curves for operation to 10 EFPY were based on an ARTwer of 110*F. 7.2.4 ARTuor_for P-T Curve Generation The ARTwor to be used for generating new P-T curves shall be selected from the plot in Figure 5. This plot is the ARTuor'as a function of EFPY for plate G-8-6 and is based on the reconumendations of R.G. 1.99 Rev. 2. While there are uncertainties with respect to the strict application of Draft Rev. 2 to BWR's operating in a low flux / fluence regime, we consider its use appropriate and conservative-to ensure freedom from brittle fracture concerns. Additionally, the t technical conuounity (i.e., ASIM, MPC, etc.) is in agreement that Cu ar.d Ni are the driving elements for irradiation-induced 1605H ,,,w..,. w., -.r-m-.. .-r e-
i TDR 725 Rev. 0 Page 24 ) embrittlement. R.C. 1.99 Rev. 1 utilises Cu and P contents for .l determining embrittlement. ] 10CFR50 Appendix G Evaluation I 7.2.5 ARTunt l The Oyster Creek operating license expires in December, 2004. Using a capacity factor of 0.8.for operation between November 1984 and ) December 2004, the additional fluecce at the 1/4T location is predicted to be 7335 days x 1930 MW x 0.8 x 1.067 x 10 n/ce*/ MWD = 1.21 x } i 10 ' n/cm*. 8 Adding this to the 0.61 x,10 mien
- received through Cycle 9 the fluence at 1/4T at the expiration of the license is 1.82 x 10
n/cm* (approximately 24.5 EFPY). The predicted ARTuor for this fluence is 142*F. i This predicted shift is less than the.200*F maximum shift permitted in 10CFR50 AppendiY. G. Should the shift exceed 200*F before end of life, the vessel beltline material must be annealed to recover toughness or additional NDE fracture toughness testing, and fracture mechanics analyses must be performed to justify further operation. Based on the results of testing the Capsule No. 2 specimens and using the predictive methods identified in-this report, the 200*F limit ( would not be reached until 50 ETPY of operation. [ l 1605H c 1, v .,..e...-, 4 ~,*-t +- - - - - - - - + - - ~ ~ - ---""-I'
- * " ^ " * ' ' ' * - ' " ' " ' * * " ' ' " ~ '
" - ' - - ~ " ~ ^ ~ ~ ' ' ~ ~ ^ ^
l i TDR 725 Rev. 0 Page 25 7.3 CVosa_, Predictions ) R.G. 1.99 Rev. 1 and Draft Rev. 2 both recommend predicting the l decrease in CVose using Figure 2. The percent decrease is a function of fluence and copper content. i As discussed in 6.4.2. there was only a slight decrease in CVues due to irradiatioa (.5 ft-lb). Since this decrease-is within the equipment's accuracy band (2 0 percent),.it is concluded tnat 5 irradiation of the capsule specimens caused essentially no decrease l in the CVvan during the exposure period. However, this cannot be assumed to remain the case for future operation. Testing and I analysis of the next surveillance capsule is required before a s judgment can be made. The plate exhibiting the lowest upper shelf energy is G-8-7 with a CVoss of 78 ft-lbs. Using a copper content of 0.20% and an EOL 2 fluence of 2.38 x 10 ' n/cm* (E > 1.0 MeV), the predicted i decrease would be 21% (16.5 ft-lb) for 32 EFPY CVose of 61.5 ~ ft-lb. This indicates that the upper shelf energy will comply with 10CFR50 Appendix G requirements at 32 EFPY. t 1605F l i
I TDR 725 Rev. O i Page 26 8.0 (;WCLUSIONS \\ 8.1 The measured shif t in the 30 f t-lb transition temperature of the vessel plate material was 72'F. The 30 ft-lb temperature of the plate material was 87.5'F which exceeded those of the weld and RAZ i materials by 60' and 55'F, respectively. It is expected that the plate will exhibit the highest RTwer through EOL. i 8.2 Tne measured shift was greater than predicted using both Rev. 1 and Draft Rev. 2 of R.G. 1.99 as shown below: Predicted Measured Rev. 1 Draft Rev. 2 72'F 40'F 28.5 sean 59.0 with margin term j Reasons for the differences may include 1. Flux effects. 2. The Oyster Creek data are outside the range of correlations. 3. Operating tem;crature effects. 4. Thernal neutron eff2 cts. S. Matsrial variability. + I-l 8.3 Two of the six beltline plates wace not in compliance with the criginal vessel specification requirements.for chemistry.. The nickv1 content rsquired was 0.40-0.70%; the measured content was 0.107%. The tcssonk for these plates'.being accepted and used are unknoww. This discrepancy has no technical effect on the integrity of the. reactor vessel. a 9 1605H ...,-,w, ..i. r
q 1 TDR 725 Rev. 0 Page 27 i l 8.4 The surveillance capsule specimens were removed from one of the plates containing 0.107% Ni. Estimates of ARTwot of the limiting plate had to include compensation for the chemistry differences. i ' I 8.5 The ARTwor for various EFPY of operation were estimated based on i the general guidelines of Draft Rev. 2 of R.G. 1.99. While it is not ] an officially released document, its use was predicated on the fact t that the technical community now recognises Cu and Ni as the controlling elements for neutron embrittlement as opposed to Cu and P. The predicted RTwot at EOL is less than the maximum allowed by 10CFR50 Appendix G. 8.6 There was no measurable effect of neutron irradiation on the CVues l of the plate material. All materials exhibited a CVoss well above i the minimum required 50 ft-lb. It is not expected that the CVess will fall below 50 ft-lbs during the remaining life of the vessel. 8.7 The change in tensile properties due to irradiation were minimal. There was a slight increase in strength and no measurable decrease in ductility. 9.0 FUTURE ACTIONS 9.1 P-T Curve Preparation and Submittal P-T curves for operation beyond 10EFPY will be generated based on the ARTwor's identified in Section 7. The new curves will be submitted to NRC by 2/28/86 as part of an Oyster Creek Technical a 160$H l __.-__--______.__._____.___________,.-s m .-.w., -..,,-.y r---. 9 -,,..
i TDR 725 s Rev. 0 Page 28 Specification change. The operating period for which the new curves t apply will be identified in that submittal. 9.2 Surveillance Capsule Reinstallation V A new surveillance capsule will be fabricated and installed in the reactor vessel during the Cycle 12R outage. The capsule will contain Charpy specimens from Capsules 1 and 2, thermal monitors, and dosimetry wires (new and those from Capsules 1 and 2). The radial position of the capsule with respect to the core may be subject to change based on the then-current knowledge of fluence and flux effects. While it may be desirable to move the capsule closer to the core to obtain EOL fluence levels before the vessel wall does, flux effects may cause changes in properties of the capsule specimens that may not be representative of what is occurring in the vessel wall. 9.3 Surveillance Capsule Testing The capsule installed during the 12R outage will be removed for 4 testing and analysis after an additional exposure time of approximately SEFPY. The exact time at which the capsule will be removed will be determined based on the progress and result of both i research and other BWR surveillance testing. GPUN will participate in and follow the progress of various organisations (e.g., MPC, ASTM, { BWROG, etc.) in more clearly understanding and defining the. effects of irradiation on reactor vessel beltline steels in BWR's. W f 1605H 4 v----e ...._,.~ ,-.9%- ,,m, r
i TDR 725 i Rev. O Page 29
10.0 REFERENCES
) 1. 10CFR50 Appendix G: Fracture Toughness Requirements 2. 10CFR50 Appendix H: Reactor Vessel Material Surveillance. Program Requirements 3. USNRC Regulatory Guide 1.99 Rev. 1 (April 1977): Effects of l Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials 4. USNRC NUREG-0569: Evaluation of the Integrity of SEP Reactor Vessels, December 1979 5. ASTM E23-82: Standard Methods for Notched Bar Impact Testing of Metallic Materials 6. ASTM E8-81: Standard Methods of Tension Testing of Metallic Materials I 7. ASTM E21-79: Recommended Practice for Elevated Temperature Tests of Metallic Materials 8. JCP&L Summary Technical Report, Fast Neutron Flux Dosimeter { Evaluation January 1974. l 9. JCP&L (I.R. Finf rock, Jr.) to USNkC (G. Lear) Letter EAGS-503, l Augus t 16, 1977 10. JCP&L (I.R. Finf rock, Jr.) to USNRC (Director, NRR), Letter dated January 12, 1978. 1605H e .._-,w..,, .m_, ,y
i TDR 725 Rev. 0 Page 30 4 l 11,0 TABLES j i 1. Beltline Material Identification i 2. Surveillance Capsule Contents t 3. Capsule Flux and Fluence Results 4 Flux and Fluence at 223.5* Asimuth 1 5. Irradiated Plate Charpy Results 6. Irradiated Weld Charpy Results ~ 7. Irradiated HA2 Charpy Results ) { 8. Irradiated Tensile Test Results 9. Limiting Plate 32 EFPY RTwor P l l l '6 l i e o t e 160$H w-- m-m.__ --,- ._,,,m_,m. ,.p. 3 .r.. w...,,, -..,.
TDR 725 I Rev. O Page 31 i TABLE 1 Material Identification Plate Code No. Heat No. ~ G-8-7 P-2161 G-8-8 P-2136 G-8-6 P-2150 G-307-1 T-1937 G-308-1 T-1937 G-307-5 P-2036 Weld-Tiller Metal Electrode Flux Weld No. Type Heat No. Lot No. i i 2-564A Raco #3 860548 4E5F 8018 HACD N.A. 8018 JBGD N.A. ) l 2-564B Raco #3 86054B 4E5F 8018 JBGD N.A. 8018 HADD N.A. l 2-564C Raco #3 86054B 4E5F 8018 HADD N.A. 2-564D Raco #3 86054B 4D4F 8018 HACD N.A. 2-564E Raco #3 86054B 4D4F 2-564F Raco #3 860548 4D4F I 3-564 Raco #3 1248 4M2F 8018 A0GE N.A. 8018 HBAE N.A. I 8 4 1605H J e e ,,w y
TDR 72$ Rev. 0 Page 32 TABLE 2 Surveillance Capsule Contents Material Specimens / Capsule Tensile Charcy Number Location Base Weld BAZ Base Weld RAZ 1 30' 3 2. 3 12 12 12 2 210' 3 3 2 9 9 9 3 300' 2 2 2 8 8 8 Note: Capsule Number 2 also contained two tensile and nine Charpy speci-mens identified as " APED Standards" (APED was the Atomic Power Equipment Division of GE). Dosimetry Wires Capsule Number Fe Ni Cu 1 3 3 3 2 3 3 3 3 2 2 2 Note: A dosimetry capsule containing three iron and three copper wires was attached to Capsule W aber 1. 1605R
1 roa 725 Rev. 0 Page 33 Tall.E 3 Capsule Flux and Fluence Results i Dosimeter Full Power Flux Fluence Enermy Material (n/cm*/see x 10') (n/ca" x 10) 1 > 0.1 MeV Fe 5.27 1.35 l Fe 4.57 1.17 Fe 4.52 1.16-Avg. 4.79 1.23 Cu 5.02 1.52 Cu 5.27 1.36 Cu 5.35 1.38 Avg. 5.51 1.42 Fe & Cu Avg. 5.15 1.32 > 1.0 MeV Fe 2.97 0.764 Fe 2.57 0.662 Fe 2.55 0.655 Avg. 2.76 0.694 Cu 3.34 0.858 Cu 2.97 0.764 Cu 3.02 0.776 Avg. 3.11 0.799 l Fe & Cu Avs. 2.90 0.746 i 1 l - i l l l 160511 4 h 6 .~
I TDR 725 Rev. 0 Page 34 TAB 1.E 4 F_1ux and Fluence at 223.5' Asinuth 1 Flux Fluence Eneray 1.oestion (n/ca'/sec x 10') (n/ca' x 10**) > 0.1 MeV Surface 7.65 1.97 AT 6.61 1.70 AT 2.95 0.76 = capsule 5.13 1.32 > 1.0 MeV Surface 3.81 0.98A (T 2.36 W. 0.61 (T 0.63 0.14 Capsule 2.90 0.75 1605H
1 TDR 725 Rev. 0 Page 35 TABLE 5 Charpy V-Notch Impact Results for Irradiated Base Metal Specimens from the Oyster Creek 210-Degree Surveillance Capsule Test Impact lateral Fracture Specimen Temperature Energy. Expansion, Appearance, Identification F _ft-lb ails Percent Shear { 122 0 4.7 3.4 0 125 40 10.5 12.2 to 60 20 .22.2 20 121 82 33.0 30,7 40 126 120 43.0 40.8 50 123 160 48.0 46.8 60 200 72.5 65.8 97 240 94.5 79,0 100 8 320 80 74.6 100 j j l 4 1605H 9
TDR 725 Rev. 0 Page 36 TABLE 6 Charpy V-Notch Impact Results for Irradiated Weld Metal Specimens f rom the Oyster Creek 210-Degree Surveillance Caosule Test Impact Lateral Fracture Specimen Temperature
- Energy, Expansion, Appearance, Identification F
ft-lb alls-Percent Shear ~ 1E2 0 9.0 8.0 10 1E4 40 30.5 26.2 35 1El 82 44.2 38.6 45 1M5 100 60.0 56.0 70 1E7 120 39.5 39.0 40 1E3 160 64.5 59.4 90 1E5 240 75.5 69.8 100 1MK 318 86.0 80.2 100 'LB 360 90.0 84.8 100 l S 1605H
TDR 725 Rev. 0 Page 37 TABLE 7 Charpy V-Notch Impact Results for Irradiated IAZ Metal Specimens from the Oyster Creek 210-Degree Surveillance Caosule Test Impact lateral Fracture Specimen Temperature Energy. Expansion, Appearance, Identification F ft-lb mils . Percent Shear 223 0 10.8 10.2 15 22A 20 18.0 18.8 '20 225 40 41.0 38.6 45 227 60 42.0 40.0 50 222 82 82.5 68.6 100 22B 100 62.0 60.0 75 224 160 69.5 64.0 100 226 240 75.0 67.4 100 23M 320 92.0 85.6 100 4 1605H
1 l TABLE 8 Irradiated Tensile Test Results L True Test Fracture Fracture Reduction Specimen Material Temp.,
- Load, Strength, psi
- Stress, in Area, Elongation Percent No.
Type F lb, Yield Ultimate Fracture psi Fercent Uniform Total 2D5 Base 150 2820 60,020 82,360 56,990 157,280 63.8 17.5 (a) 31.1 (a) 2D6 Base 230 2680 58,590 79,700 54,140 169,940 68.1 16.5 (a) 29.5 (a) 2D4 Base 550 3230 57,640 83.670 65,440 146,750 55.4 13.0 (a) 22.2 (a) Unirradiated Base RT NR 64,000 84,900 NR NR 66.9 NR 27.5 (b) 2L4 Weld 80 3150 73,520 92,740 64,060 176.180 36.6 10.9 (a) 23.1 (a) 2LL Weld 240 2940 71,730 87,970 59,320 173,250 65.8 15.0 (a) 27.5 (a) 2L3 Weld 550 3350 71,170 92,360 67,920 154,590 56.1 9.6 (a) 21.5 (a) Uni rradia ted Weld RT NR 65,000 84,000 NR NR 67.0 27.5 (b) 2UU HAZ 82-3050 70,290 90,210 61.340 178,990 65.7 10.5 (a) 20.9 (a) 2TC. BAZ 550 3450 65,560 89,960 69,280 150,720 54.0 10.0 (a) 19.2 (a) (c)- The elongation is for a cae-inch gauge length. 2's (b) The elongation is for a two-inch gauge length I.* d NR = Not r,eported. . u o U, ou u
YDR 725 Rev. 0 Page 39 v TABLE 9 l l Limitina Plate 32 EFPY RTwor(*F) i R.G. 1.99 Rev. 1 Evaluation i 32 EFFY Plate CF ARTwor + RTwer
- RTuor,
= G-307-1 148 72 + (+15) 87_ = G-308-1 148 72 + (+15.5) 87.5 = i l G-307-5 215 105 + (-19) 86 = G-8-6 185 90 + (+12) 102 = .l G-8-7 215 105 + ( -4) 101 = G-8-8 160 78 + ( +5) 83 = s R.G. 1.99 Draft Rev. 2 Evaluation l 32 EFFY Plate CF ARTuor + RTwor RTwor = G-307-1 79 48 + (+15) 63 = l G-308-1 79 48 + (+15.5) 63.5 = G-307-5 141 86 + (-19) 67 = G-8-6 138 84 + (+12) 96 = G-8-7 135 83 + ( -4) 79- = G-8-8 132 81 + ( +5) 86 = i 4 ) 1605H y + w
l 1 TDR 785 i Rev. 0 l Page 40 i 12.0 FIGURES 1. Vessel Beltline Material Location 2. Irradiated Base Metal Transition Curve 3. Irradiated Weld Metal Transition Curve i t 4 Irradiated EAZ Transition Curve e 5. ARTuor vs. ETPY i i f 1 l 9 e l l l 1605H l l s
wn us Rev. 0 Pcg3 41 g' i g.s.s 2 sse l \\ Lawn G-t
- Iutsamtoints
/ g teLLL TDP OF 4 ACTivt Futb \\ 270* Aebuta. i e. 1,. i lit 2-54YA l 26 [ C M$vLE G.g-t l 1 s.acenau ac* ll j i i M 3-st.1 p2 56YE g.. l '[ i G.%] 5 q.301.( I 80 Tram en I 1 I- ~ ~ SQ* ACTuvs vasc L10' ' - - ~ ~ 2-St3D " ( ' ] 2-S4F s 1 g.g.g 1 /so* FIGURE 1 1 l
TDR 125-Rev. 0-i .P:33 42 s. s 4 ' ) 0 l g. r ?. t i i g. .l ,..F g,/ / 1 +
- o
- i l
l~ l o l. V l !.:j j l l \\ ;. m ll l 18 ll l l \\ l l \\ ,l l l w l l Ls,. l l l l l l b i l$ l l l OC .b .i $~ 0l lC l W l l l l E O. .o p l. j o: f,) o p' Q. MedletM ./ O treenleted l ,/ ./ - - Welhus Fit ~ j/,f' / g; a SS Confidense 1.imRs l 0/ ./ a -se o-so no iso zoo 2so soo sao 4co TEST TEMP, F r i P l CHARPY V. NOTCH IMPACT ENERGY VERSUS TEST TEMPERATURE l-FOR THE IRRADIATED ~8ASE METAL SPECIMENS FROM THE ~ OYSTER CREEK 210-DEGREE SURVEILLANCE CAPSULE - 1 s l;' 5 FIGURE 2 l + t a r i. I es > y.? Y g_._.y,._,,..,.m
m-+
. 1. .,,,_...,,..._,_.g.,,,,my ,,.o%., ,,y. c.
_... ~. TDR 725- ! A v. 0 1 Pese 43-O Wredleted l Weibut Fit 95 Canfidence Limits 8-3. e o 8 e a a a *o. a g l s V w 8 >gf O g / 0 g g. / ,./ a i W / O Unf eradiated Baseline / (Average Values for i g a f' Various Flux-Lots) .~ N l / 8-a g W Y E I ~ g g g g g W W V g g 7 I g -so-o so-1oo iso 2co 2so soo. .sso 4oo l TEST TEMP, F CHARPY V-NOTCH IMPACT ENERGY VEkSUS TEST' TEMPERATURE FOR THE IRRADIATED WELO METAL SPECIMENS FROM THE OYSTER CREEK 210-DEGREE SURVEILLANCE CAPSULE h i FIGURE 3 t 1 e s, -e -~+.h.--
TDE'125-l Rev.-0 Pcg3 44- -i t i h O Irredlet ed- - Weibull fit E; SS Confidense Limite \\ 3............................. '.,,........ \\ E: c l., a a s. a i ' S-l ...c' I n g . >. g; / / e l l a: Wa' ,.D / U w l R-l l i a n-a:: e i al o/ 8 a a i i i i i s i -50 0 50 10 0 15 0 200 250 300 350 400 i TEST TEMP, F CHARPY V-NOTCH IMPACT ENERGY VERSUS TEST TEMPERATURE FOR THE IRRADIATED HAZ METAL SPECIMENS FROM THE .0YSTER CREEK 210-DEGREE SURVEILLANCE CAPSULE FIGURE 4 p j ' 9 _________,__.__.__.m._ - - - - - - - - - - - - - - - - - - - - - - - - - - ^ - - ^ - ' ^ ^ - -
TOR 728 MEV. 0 . P80s 45 : 4 \\ l 1 1000 e e i a i i e1l 1 1 6 iiaiL ) 4 t 1 500 le I I \\ t i-i ARTNOT l (9x10') 100 'I i i 50 ~ i l l ' ' ' ' ~ 'l' I I I I 't i 10-1 5 10 80 100 EFPY (x 10*) l Figure 5 - ARTNOT
- W I
l I I i 1 G e .......c.-, ...+....---------*--d-------'"--*------'----~
TDR 725-
- RGv. 0 :
Page 46 APPENDIX'A Unitradiated Material Data Appendix A.1.- Plate Appendix A.2 - Weld i l I'j l i t q 0 1605H e --.-.A-------------------------------'- ' - - ' ^ ^ - ' ^ ' - ^ - ^ ' ' ' ' ' ^
A TDR:725-Rov. 0. . Page 47 I APPENDIX A.1 Plate t4aterial / 4 p'. 1 -1605H -5
TDR 725- -RLv. 0- .l .Page 48-CHEMISTRY (*) ELEMENTS Code No. Heat No. C ~ Mn P S __ $1 Ni-Mo 1 G-8-7 P-2161 0.19 1.35 0.019.~ 0.021 .0.24 0.48- .0.46 i G-8-8 P-2136 0.19 '1.36 .0.006 0.024 0.26' '0.46-0.48 O G-8-6 P-2150 0.2-1.25 0.013 .0.026 0.23 0.51 -0.46 G 307-1 T-1937 0.2 1.4 0.011 0.022 0.24 NR 0.51? r G-308-1 T-1937 0.2 1.4 0.011 'O.022 0.24 NR 0.51-t G-307-5 1?-20% 0.2 1.28 0.019 0.030 0.21 0.53 0.'52
- REPORTED by Lakenr. Steel l
NR = Not reporced TENSILE PROPERTIES (*) i Ultimate Elongation. ~ Reduction Code No. Heat No. Tensile (ksi)- Yield (ksi) in'2" (1) of Area % i G-8-7 P-2161 88.6 67.2 25.0 66.0 i G-8-8 P-2136 90.0 .69.0 24.0 66.3 i G-8-6 P-2150 85.1 62.7 28.0 66.0-G-307-1 T-1937-2 86.7 65.0 ,26.5 66.2 G-308-1 T-1937-1 84.9 64.0 27.5. 66.9 G-307-5 P-2036 88.5 66.5 28.0 .66.4:
- Performed by C-E following plate heat treatment of 1550'-1650'T for 4 hours, quench, 1225'! 25'F' for 4 hours,1150 25'F for 30 hours, furnace cool to 600'F.
j, I 1. 1605H i-
p.4 TDR 725 i. ( 6 v ,t - t -,- -. Rov. 0 t.. Page 49 i t 4 i i i - - -..1. ~.. - _1_. t p-_ t- __t. .-i n + 3-i j p.
- a.
.. _ _..,. - --t r--~~ -1 ,,g- --I-i. t.-- .r. p. I,+.. - -.r,a r.,.r i,,- -m,,-l ,, e r +;. . ;+ ;- - W pp__q q l u_ _.- a.a... 'l .p:~._. _ ;.. _ _. u.-... _1.._._.. .... l
- _._,-j.-._-_t..-.-
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l L TDR 785-1 I Rev. 0 l Page 57-l Charpy Data (*) Wire Heat Flux Lot Test Impact No. No. Temperature ('F) Enerry (ft-lbs) W5214 SG13F +10 ~ 61 52, 58 (57 Avg. ) 860548 .4E5F +10 66, 64.5, 65 (65 Avg.) 86054B 4D4F +10' Not Available 1248 4M2F +10 53.5, 57.-65 (58.5 Avg.) Note: There is no data available for 8018 electrodes. Performed by C-E following heat treatment of 1150'+t$'F for 40 hours, furnace cool to 600*T. l . ) 1 I e 1605H i ____.__.-__.--------------------L----'-------^- - ^ ^ - - ^ ^ ^ ^ - - ^ ^ ^ ^ ' ^ ^ '
- TDR 725 g,j, R:v. 0 Page 58 4 APPENDIX B 3attelle Columbus Report CS Surveillance, Material Testina-l- k i e C p 5 I n 6 h I 1 t } i ' 4 1605H I '. p n .n.. n m s--n . --b
ATTACilMENT.4 713t 725-Rev. 1 Page 12 ' 7. 0 ' gg and Ch. MEDICTICMS 7.1 Int 2ctkaction ? To develop P-T curves for an operatirg period, the adjustad Eg at the ord of that period nust be conservatively and reancmably estimated. This section describes the asthods used to predict the' adjustad E , and the results, for several operating parieds including f-license (EOL). Also, the offacts of-irradiation on the oystar Creek beltline materials' CVUSE are assessed. 7.2 g g etions + 7.2.1 Predictive Methodoloav (R.G. 1.99 Rev. 2) 7.2.1.1 Plate Material The plate materials'ocpper and nickel ocntants are known. The fluence factor is to be cala21ated using the surface r fluence values provided in Ref. 14. The ratio of @ e at kT l-to %e at the surface is obtained frtan the Battalle e i. provided in Appendix B. The initial Eg-is obtained fztza the transition carves provided in Appendix A.1. The standard deviation.( I) is.to be based tpon the precision of the' test method used' to determine IE ASD(/E-23-86, the Omrpy testing specification,- has rc g. l precision yet established. Section 13 states that-precision and accuracy of these authods are being established. Section 10, idtidt specifies the verification .I requirements for test equipment, requires that equipment obtain absorbed energy values within 1 ft-lb or 54, idtidsver is greater, when standardized specimens are testad. Within the expected energy range (30-50 ft-lbs) at C temperatures 60 F above the E ft-lb differenosisinsignificantasYar,a1-2 l as temperature-is careerned. Mcwever, for conservatism, we reviewed the l plate transition curves and estimated a marin = 5 F 0 i difference between Tcv30 and'Tcv31.5 (30 ft-lho + 54). i 1herefore, a I value of S F will be used in the plata C notarial caloalations. 7.2.1.2 Wald Materig ~ 4 Pbr the otsubination of 1248/4M2F, the copper and nickel contents frtza the EPRI data base (Raf.12) shall be used. For the ocabinations of.86054W4D4F and 86054W4E5F, a copper ocntent of 0.35% will be used as specified in R.G. I 1.99 Rev. 2, section 1.1. i j 4 4
TtR 725 Rev. 1 Page 13 We anticipate that this will patovide very conservative remits since the as-deposited copper content of the - tandesparc deposit in the EPRI data base was 0.22%. ocrisidering that it is hi$ly rMle that both wires were ocyper contad and dilution coeurred, we consider that is is unlikely that the as-deposited copper content of the-86054W4D4F and 86054W4E5F. is near the 0.35% used in the caloalations. The nickal content for.86054W4D4F and 86054 W4E5F will be 0.2%. We consider this to be a conservative upper bound 1 value even thouth the R.G. rummunards that we use a value of 1.0% if no data is.available. - The as-deposited nickel I contant of the tandenrarc weld was 0.046% whie is consistant with an expected value for weld wire with no nickel purposely added.: 1he nickel contant of the 1248 ) heat.of wire in the EPRI data base is 0.11%, and the nickel; r contant of the surveillance weld wire heat-is 0.054. Therefore, the use of 0.2%'for the 86054B heat of weld wire is considered conservative. Ths fhw ra factor is to be calaalated in the sans manner au 6m for the plate. l IRT is to be specified as described herein For wire l nmb48, a value of -31 F is to be used. 0 This is the Tcv30 of this heat as shcun in the EPRI data base. For l wire heat 860548,'a value of -8'F will be used. This value-is mch hi@er than the -56' value to be assumed for Arcos B5 flux welds specified in the PIB rule-(Ref.10). l However, the Charpy test results~ for 86054W4D4F were lower at +10'F than for the 86054W4E5F. and the other two caihinations as= listed in Appendix A.2. We arrived at the -8'F value by using the GE estimation method provided in l Ref 11. Using this method on the other a:mbinations, we. obtained a value of -57'F which is very close to.the value-specified in the PIS rule. Therefore, we consider that using an estimated IRPNDP of -8'F is W ale and ocnservative. For U I, we use a value of 5'F for. wire heat 1248 since it ^ is a mmannsd value.. For wire heat 860548, we use.a value j. of 28'F. !?ds value of I is equal to the value of =. which la tes the margin to be applied to the mean shift tanparature.. We consider this to be a cxmeervative we since the -8'F value W is based on available-data and is auch higher then the mean IRrNDP value used in t2w, FIS rule.- f s i s
- *~+-
r e = ee e. w e e--e.. .-,em ee -- -e ss, w -e-p g -w + v, y-w v gw.,.--r-,,ve-w,Yi-- t vw-ww ty wg ve, w e * -4 ** ---
^ i 'IER 725 i 7 Rev. - 1 Page 14 7.2.2 Limitin2 Material Selection of the limiting matarial (i.e., the natarial Wiich will exhibit the highest RF at the end of the operating period) depends on two actors: - 1) initial urtirradiated RT mterial chmaistry. 'Ihe adjustaant to M and 2) to RT is defined in 10CfftSO Appendix G as "the tanpara shift numanrad at the > 30 ft-lb (41J) level in the average Charpy curve for the. irradiated material rulative to that for the unirradiated-mtarial". Selection of the limiting notarial will focus 'an both the plate and wald materials. Front 7.2.1, the initial unirradiated RT and the: essential chenistry contant (Cu and Ni For R.G.1. ii.v. 2 evaluation) are summarized below: Plata Initial cada _R_rg 02 Ni G-308-1* +15.5 0.173 0.107 1 1> 1 >Sama Heat G-307-1 +15 0.1731 0.107 (T-1937) G-307-5 -19 0.27 0.53 G-8-6 +12 0.20 0.51 G-8-7 -4 0.21 0.48-G-8-8 +5 0.18 0.46 Wald Material 86054B/4E5F. -8 0.35 0.20 86054B/4D4F -8 0.35 0.20 1248/4M2F -31 0.22 0.11
- Surveillance capsula plata.
1Surveillance capsule results. An evaluati ofh(RT at a 15 EPPY surface fluenos E N MeV) was performed for eacts of.1.75 x 1 8 W plate and wald (Ref. 13). 'lhe'results are provided in. Table 9. The limiting material is weld wire heat number 860548. 7.2.3 ARP Predictions for the U=4 ina M!tterial t 'Ihe ARP for the limiting material calculated usirq the methods identified in RG 1.99 Rsv. 2 are tabulated below for the identified operating period: i 1
TTR 725 Rev. 1 Page 15 l Surface Fluence Operating (E > pV) 18) Period (EFPY) CF (Was x 10 g 8.38 259 0.98 105 15 259 1.75 129 32 259 3.74 173 7.2.4' M NDT for M Curve Generati_cn l The ARr j 15 E d 129'F.to be used for generatirq P-T curves through The ARP used for the existing curves was l 125'F. i j 7.2.5 Rrg - 10CFR50 Accendiv G Evaluation I T'ls predicted shift is less than the 200'F mavi= P shift { perndttal in 10CFR50 Appendix G. Should the shift exceed 200*F before and of life, the vessel beltline material aust be amenled to recover toughness or additional NEE, fracture tauginess testing, and fracture mechanics analysee nust be performed to justify further operation. Based on the calculation (13), the shift is 102*F at 32 EFPY, well below the 200*F =avi= = allowed. 7.3 _CV,w, Predictions R.G. 1.99 Rev. 2 zurw=marids predicting the decrease in CVUSE usirq Figure 2. The percent decrease is a fury:: tion of fluence ans copper ocntent. As dier = =ad in 6.4.2, there was only a slight decrease in C Q ( due to irradiation (.5 ft-lb). Since this decrease is within tne equipnent's accuracy band (_+5.0 percent), it is concluded that irradiation of the capsule specimens caused essentially no decrease in the CVUSE during the exposure period. However, this camot be a===ad to remain the caos for future operation. Testing and analysis of the next surveillance capsule is required before a judgment can be ande.- 1he plate euhibiting the lowest upper shelf energy is G-8-7,with a l CVg of 78 ft-lhe. Using tg m~ (E > 1.0 MeV),' the content of 0.21% and an' EDL l surrace fluence of 3.74 x 10 year predi* ad decrease would be 16.5 ft-lb for 32 a EFPY CV ggg of 61.5 ft-lb. 'Ihis indicates that the upper shelf energy will cumply with l 10CER50 Appendix G requirements at 32 EFPY. l i-1 ,-c.. ,.,,-u.,.,
.= IIR 725 Rev. 1 Page 16 8.0 CENC21 bit l35 8.1 The measured shift in the 30 ft-lb transition tauperature of the vessel plate material was 72*F. The 30 ft-lb tanparature of the plate material was 87.5'F which 9medmi thces of the Wald and HAZ materials by 60' and 55'F, s&ively. 8.2 The maasured shift was greater than predicted using Rev. 2 of R.G. 1.99 as shown below: i e predicted 72*F 29'F mean naasons for the differences may includa: 1. Flux effects. l g 2. The Oyster Creek data are outside the rarge of corzulations. 3. Operating tenparature effects. 4. Thermal neutrun effects. 5. Matarial variability. 8.3 Two of the six beltline plates were not in otsqpliance with the original vessel specificaticn requia.uum-d for cba=4=try. The nickal centent required was 0.40-0.70%; the measured content was 0.107%. The reasons for these plates' being av.mpani and used are unknown. This di m % cy has no technical effect an the integrity of the reactor vessel. 8.4 The surveillanos capsule specisam were removed from one of the plates ocntainirg 0.107% Ni. Estimates of RF of the limitirq material had to include ocupensation for the differences. I 8.5 The RT for varicus EFPY of operaticn were estimated based cn the qui inas of Rev. 2 of R.G.1.99. The predicted maan shift RTygyp at EOL is less than the marintan allowa$ by 10CFR50 Appendix G. 8.6 there was no measurable effect of neutztn irradiation en the CV } m l of the plate ratarial. All materials exhibited a wall above the minimum required 50 ft-Ib. It is not expected the CV will fall below 50 ft-lbs durirg the remaining life of the 8.7 The change in tanelle r W ies due to irradiation ween minimal. There was a slight irus in ei.amap.h and no measurable decrease in ductility. l
TER 725 Rev. 1 a Page 17 9.0 FUIWE AC22@ 9.1 &Irveillance daa=He Rainstallation I A new sutveillance r apmile will be fabricated 'ard -installed in the reactor vessel during the cycle 13R ca.rtage. The capsula will i cantain dbarpy specimens from Capsules'1 and 2, thermal monitors, and dominstry wires (now,and those from Capsules 1 and 2). The-radial position of the capsule with respect to the oors may be subject to change based on the then-current knowledge of fluence and flux effects. While it may be desirable to move the capsule closer to the core to obtain BOL fluence levels before the vessel wall does, flux effects may cause changes in p.vy.mi.ies of the capsule specimens that may not be representative of Wat is occurring in the vessel all. 9.2 Surveillance omr=He Testim The capsule installed during the 13R outage will be removed for tasting and analysis after an additional expceure tima of. -cye dmately 5 EFPt. The exact time at Wnich the capsule will' be-removed will be determined based cri the progress and result of both raaaarch and other BIR surveillanos testing. GPUN will participata 1 ~in and follow the progress of various organizations (e.g., MFC, ASDE, BIROG, etc.). in more clearly urab..umding and defining the. 1 effects of irradiation en reactor vessel beltline steals in ENR's. 10.0 REFERDRES 1. 10CFR50 Appniix G: Fracture Ibuq$ ness Requi._ tem 2. 10CFR50 Appendix H: Reactor vomaal Material Surveillance F.w.m Recpirements 9 3. USNBC Regulatory Guide 1.99 Rev. 2 (May 1988): Effects of Residual Elements on Predicted Radiation Demage to Reactor Venami Materials 4. USNRC NURB>0569: Evaluation of the Integrity of SEP Reactor Vessels, narva*we-1979 5. ASDE E23-82: Standard Methods for Notched Bar Impact Testing of Mut:allie lentarials 6. ABBt ES-61: Standard Matheds of Tension 'Nating of Metallic Materials 7. ASTM E21-79: Recommended Practica for Elevated Te tdre Tests of. Metallic Materials - 8. JCP&L Samary Technical Report, Fast Neutron Flux Dosiaster Evaluation, Jarnaary 1974.' g 9. JQ%L (I.R. Fir awG,. Jr.) to USNRC-(Dizinh, NRR), Letter dated Jamary 12, 1978.
TIR 725 pay,.1 l. Page.18
- 10. 10GR Part 50.61, Fract:ure 'Ibu@ ness Requirements for Probaction Against Pressurised '1hernal Shock Events.
HEDC-31140, IBR owners' Group Evaluation of Regulatory Guide 1.99, 11. Revision 2 Iapact on BNRs, GE, Jarnary 1986. 12. EPRI NP-4797, Nuclear Plant Irradiated Steel Handbook, Spanhar 1986. 13. GRE Calculation No. C-1302-221-5340-010,Rev. O. 14. GPUN Mano 5411-88-0168 (10-13-88): Updata of 14anctor Vesaal Fluence Projections. i - i l 9 C e
.-6_ - o I TER 725 HUV. 1 Page 28 TMEE 9 - M Pralictions at 15 EFPt Plate g g.7) G-308-1 73 G-307-1 72 G-307-5 77 G-8-6 95 G-8-7 80 G-8-8 82 Walds 86054B/4E5F -129 86054B/4D4F 129 - 1248/4M2F 43 t .)
d@ AIO ///// Mkg ?;. IMAGE EVALUATION N/ TEST TARGET (MT-3) 1.0 lf m La y@Its l-l E m En t: 1.25 1.4 j l.6 l 4 150mm 6" $>%&A + + a,, 4 'b' 7/ g$+$ PHOTOGRAPHIC SCIENCES CORPORATION 4)4% P.O. BOX 338 WEBSTER, NEW YORK 14580 j (716) 265-1600
+k+ lbo ///gf\\9 t +# IMAGE EVALUATION 8 @/ k///7 4 $If TEST TARGET (MT-3)
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4# $/ II///O k [>ff// g+/ %g% [jjp// IMAGE EVALUATION ,jp/// %, TEST TARGET (MT-3) f+ 4 i 1.0 E2 M yll NL l,l bb U8 1.25 1.4 1.6 l l 4 150mm f 4 6" d *'% /a4 f4
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- p 4 f P.O. BOX 338 f\\g WEBSTER, NEW YORK 14580 4
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III///o kfI>@ ///j// 8/ t IMAG . VALUATION 4, l.0 lff 88 L'4 ,5 IW- _m D h$$ l,l l2 J.25 { i.4 i.6 4 150mm 4 6" 8?e%i /$ ++);3<p
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~ ATTACHMENT: .[ }lk .3'~ y,.
- UNITED STATES ~
TT W ',. g~ [ NUCLEAR REGULATORY COMMISSION ^5" WAssmorou. o. c. gosss y F.] : s,. }~ '56,,. ' June 28, 1983' g-q s. r-g i*
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Docket No. 50-219 t.505-33-06-063 ,) i -3Qp%. Mr. P. B.. Fiedler 9 . Vice President and Director 9 Oyster Creek Nuclear Generating Station g-Post Office Box.388 jo
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j. Forked River, New Jersey 08731 j y,, 84 Qg w Ir
DearIsr.'Fiedier:
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SUBJECT:
INSERVICE INSPECTION (ISI) REVIER N Oyster Creek Nuclear Generating Station-( By letters dated ' June' 15 and November 15,1978, June 8, September 6, and December 11,1979, June 5 and August 5,- 1981, May?2,; August 20 and ! .t September. 14,1982, you submitted the inservice inspection program, revi-sions, or additional infonnation' related' to requests for-relief:from certain Code requirements determined to be' impractical to-perform on the Oyster Creek Nuclear Generating Station, pursuant to 10 CFR'50.55a(g). Tha purpose of this letter is to inform you of the results.of-the staff review of your relief requests and to grant relief-in part from the requirements of Sec. tion' XI of H the American Society of Mechanical Engineers Boiler and Pressure Yassel Code (the Code) or impose other requirements, as appropriate. The review of requests for relief from the inservice inspection requirements has been c:moleted. Based on the results1 of. this review, the staff has : determined there are cases in which the requested relief is granted as-proposed. Section 50.55a(g) of 10 CFR Part 50 requires. that your' program be revised at-l 120-month intervals with the. start of comercial; operations being the refer-h ence date. The start of the next interval for your. facility is' Decemoer 8, 1989 and your inservice inspection:and testing, program must be based on the edition and addenda of the Code incorporated by 10-CFR 50.55a(g) 12 months prior to that date. Any changes to your Technical Specifications are recuirea j' to be submitted at least six months prior to -the beginning.of 'a.120-month. interval and it is requested that;any requests for relief from Code require- ments be provided on the same schedule. The staff review of ~your relief recuests for your next interval will be conducted on a schedule basec.on the ~ - program-revision requirements for your facility. Until. that time you -should folicw the inservice inspection program proposed and.as" described herein ano further relief granted or additional testing imposed during-the remaincer'of-F the period. Any* relief.from Code requirements granted herein ' expires c6 December 7, 1989. I qso104OS69
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- g Mr. F. B. Fiedler June 28, 1983 The enclosed Safety Evaluation, as supported by our contractor's Technical Evaluation Report (TER) attached thereto, delineates those items for which relief has-been granted and alternate schedules and procedures defined.
.The staff has determined that, where stated, the Code requirements are 'im-l _' practical, the granting of t51s relief is' authorized by law and will not endanger life or prope:-ty c; e.: enninon defense and security, and is other-i wise in the public interest considering the burden that could result if they j were imposed on your facility. Jour.peoposed. Technical Specification changess .l ~ related to this ' issue will be. handled'in.r separate action.1 1 A copy of the related Notice of Granting of Relief is enclosed. Sincerely, k Dennis.M. Cru e d,/JChief. Operating' Reactors Brdch #5 Division of Licensing
Enclosures:
1. Safety Evaluation Report l with Technical Evaluation Report prepared by Science Applications, Inc. j 2. Notice 4 cc w/ enclosures: See next page -l 1 t I l i t 1 -i 1 'k I .i I O
p,- t e l Mr. P. B. Fiedler June 28, 1983-cc G. F. Trowbridge, Esquire Resident Inspector - 1 Shaw, Pittman, Potts and Troweridge e/o V. S. NRC 1800 M Street, N. W. Post Office Box 445 Washington, 0. C. 20036 Forked River, New Jersey 08731 -l J. B. Lieberman, Esquire Comissioner Berlack, Israels &~ Lieperman New Jersey Department of Energy .i 26 Br.oadway 101 Comerce Street, New Ycrk, New York 10004 Newark, New Jersey 07102 Or. Thomas E. Murley, Frank Cosolito, Acting Chief Regional, Administrator Bureau of Radiation Protection - Nuclear Regulatory Comission, Region I Department of Environmental 631 Park Avenue Protection King of Prussia, Pennsylvania 19406-380 Scotch Road Trenton,- New Jersey ~ 08628~ i J. Knubel BWR Licensing Manager GPU Nuclear 100 Interplace Parkway Parsippany, New Jersey 07054 Deputy Attorney General State of New Jersey Oepart.:ent of Law and Public Safety 36 West State Street - CN 112 Trenton, New Jersey 08625 Mayor-Lacey.Tcwnship 818 Lacey Road Forked River, New Jersey 08731 U. S. Environmental Protection Agency Region II Office ATTN: Regional Radiation Representative 26 Federal Plaza ,New York, New York 10007 l Licensing Supervisor i Oyster Creek Nuclear Generating Station Post _0ffice Box 388 l Forked River, New Jersey 08731 l o a
itCy , ~,[ a UNITED STATES'
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NUCLEAR REGULATORY COMMISSION i * ., ASHING TON, D. C. 20565 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION-RELATED TO REOUESTS FOR RELIEF FROM INSERVICE INSPECTION REQUIRDfENTS ] GPU NUCLEAR CORPORATION AND l i JERSEY CENTRAL POWER AND LIGHT COMPANY DOCKET NO. 50-219 INTRODUCTION Technical Specification 4.3 for the Oyster Creek Nuclear Generating Stantin states that inservice examination of ASME Code Class 1, 2, and 3 componeer.s~~ l shall be performed in accordance with Section XI of the ASME Boiler assed. - Pressure Yessel Code and applicable addenda as required by 10 CFR 50.55a(g) l except where specific written relief has been' granted by the Comissicze. Certain requirements of later editions and addenda of Section XI are ilub-practical to perform en cider plants because of the plant's design, cc:suponent j geometry, and materials cf construction. Thus,L10 CFR 50.55a(g)(6)(fi l authorizes the Comission to grant relief from those requirements upont j the necessary findings. making i q By letters dated June 15 and Novemoer 15,1978,.-June '8, September 6, amo l Decemoer 11,1979 June 5 and August 5,1981, May 2, August'2, August - 26, and Septemoer 14, 1982, General Public Utilities submitted its inservil<ce. inspection program, revisions, or additional information related to reatquests for relief from certain Coce requirements determined to be impracticall: to-perform on the Oyster Creek Nuclear Generating Station. during the.inspeection interval. The program is based on the 1974 Edition through Summer 1975 Addenda of Section XI of the ASME Code, and covers :the 120-month inspection-interval from Decemoer 8.1979 through December 7,.~1989. DISCUSSION AND EVALUATION Requests for relief from the requirements of Section XI which have beens ~ determined to be impractical to perform have been reviewed oy the Staff's contractor, Science Applications, Inc. The contractor's evaluations onf tho-licensee's requests-for relief 'and his recomendations are presented 11:n the ' Technical Evaluation Repcrt (TER) attached. The staff has reviewed ttnan TER. l~ and -agrees with the evaluations and recomendations except in the casee of = Examination Category B-J, Item No. B4.5, inaccessible welds not:in penetra-tions. It is the staff's position that the welds for which relief ~is-requested should be identified in ' order that the impracticality of them examination may be determined and the impact on safety.of not performic:ng-the examination evaluatec. A summary of the determinations made by tmee staff is presented in the following.tacles: $901D W Y b@Obb-
Lg- !s i e, l TABLE 1 CLASS 1 COMPONENTS LICENSEE PROPOSED. RELIEF IWB-2500 IWB-2500 SYSTEM OR' AREA TO BE REQUIRED ALTERNATIVE REQUEST ITEM NO. EXAM. CAT. COMPONENT EXAMINED - METHOD EXAM. STATUS Bl.1 B-A Reactor Vessel Volumetric Visual During GRANTED Vessel-Welds in Pressure Test Core Region ~Each Outage Bl. 2 B-B Reactor Circum-Volumetric Visual During GRANTED Vessel farential. Pressure Test ~ & Longi-- Each Outage tudinal Welds-in e Shell (Other Than Those in Core Region)' Bl.4 B-D-Reactor-Primary Volumetric Visual During GRANTED (1) _ Vessel Nozzle-System Pres-to-Vessel sure tests i Welds and Inside Radiused Sections: N1A N2A N1B N2B N1C N2C N1D N2D NIE N2E & N9 l Bl. 6 _B-F Reactor Recircula-Volumetric . Visual During GRANTED (1): Vessel tion Nozzle - and System Pres-to-Safe End Surface sure Test Welds: Each Refuel-NG-A 'ing Outage NG-B-25 and Each-- NG-C-24 -Interval NG-D-23 NG-E-27 NG-A-1 NG-B-1 NG-C, NG-D-1 NG-E-1 NC-4-1 NP-2-1 NP-2-2 J (1) Augmented examinations performed under I&E Bulletin No. 82-03. provide basis for relief. +
)~ ~ +, ) ~, TABLE 1 (CONTINUEDT I CLASS 1~ COMPONENTS LICENSEE PROPOSED RELIEF IWB-2600 IWB-2500. SYSTEM OR AREA TO BE REQUIRED ALTERNATIVE REQUEST ITEM NO. EXAM. CAT. COMPONENT EXAMINED-METHOD' EXAM. 5TATUS Bl.13 B-I-1 Closure Cladding
- 1. Visual Assess Gen-GRANTED Head-and: Sur- - eral.- Condi-faca
- tion of or Cladding
- 2. Volu-metric Bl.14 B-I-1 Reactor Cladding Visual Asses.i Gen-GRANTED Vessel eral Condi-tion of-Cladding Under B-N-1 Category Bl.S B-F Reactor Nozzle-to-Volumetri:
'- umetric GRANTED-Vessel -Safe End and
- Oma fj Welds:
. Surface
- Only g M g*
NSA arc
- 00% -
S R u.a ! N3B )r - r a N B4.5 B-J. Piping Inaccessible Volumetri'c Visual of - WITHDRAWN . Pipe Welds . Areas of in Penetra-Penetra-L tions: tions X-SA. X-6 During. l X-5B X-70 Hydrostatic X-3A' X-72 Tests. L X-3B X-7 X-2A X-8 X-2E X-9 X-4A X-10' X X-61 X-12B l B4.5 B-J Piping Inaccessible Volumetric Visual Ouring WITHDRAWN. Welds.Not In Hydrostatic Penetration Test 84.9 B-K-1 Inte-Support Volumetric Visual REQUEST grally Welds WITHDRAWN Welded 7/82' Supports t w - - - - - - -. a e-w-
. * - C:t .1 . TABLE 2 CLASS 2 COMPONENTS LICENSEE PROPOSE 0 ~ RELIEF-IWC-2600 IWC-2520 SYSTEM OR AREA TO BE REQUIRE 0 ~ ALTERNATIVE REQUEST. ITEM NO. EXAM. CAT. COMP 0NENT EXAMINE 0 METHOD ' EXAM. STATUS l C2.5 C-E-1 'Inte- ' Pipe Sup-Surface' None .WITHORAWN grally ports Welded Welds 7 Pipe for Pene-Supports trations: i X-63 I X-66 X-23. t X-20A' X-208 C2.6 C-E-2 Pipe' Support. Visual None WITHORAWN Supports Components for Pene-trations: X-63 X-66 X-23 X-20A 1 X-208 C1.1-All Portions' Welds Volumetric Visual WITHO." AWN C1.4, Cate-of Systems or-C2.1-gories (Other Surface C2.6, than ECCS) and Visual C3.1-which do C3.4, not Func- & C4.1-tion During L C4.4 Normal .l Reactor Operation .(Static Systems) l 1 l I l
.:r T ? .., i u a -TABLE 3 CLASS 3 COMPONENTS - l No Relief Requests - t L i l } 1 . i . [r - i e e l l l r t e 1. 1 ( i L I i 3 a v e
s: ,y e, TABLE 4 PRESSURE TEST IWA-5000 IWB-5000 IWC-5000 & LICENSEE PROPOSED RELIEF ~ SYSTEM OR IWD-5000 TEST ALTERNATIVE REQUEST COMPONENT PRESSURE REOUIREMENT TEST PRESSURE STATUS Class 1 Requirements of 1974 Requirements of 1980 GRANTED-2 & 3-Edition, Summer 1975 Edition, Winter.1981 Systems Addenda Addenda 9 J ~ i a l i l i i A .k i .~ 4 )
I'.. -. . y' ' ] l TABLE 5 l ULTRASONIC EXAMINATION TECHNIOUE LICENSEE PROPOSED SYSTEM OR ALTERNATIVE RELIEF-REQUEST COMPONENT REQUIREMENT EXAMINATION METH00 STATUS-Class 1 Requirements Requirements GRANTED &2 in 1974 in 1977 Edition. Systems Edition. Summer 1978 I-Summer 1975 Aodenda Addenda l l i l s e.: 4 . I e - ( ~m
~ a c 2-Based on its review, the staff concludes that relief granted from the l examination requirements and alternate methods imposed ttirough this I document give reasonable assurance of the piping and component pressure boundary and support structural integrity, that granting relief where the Code requirements are impractical is authorized by law and will not endanger life or property, or the connon defense and security, and is otherwise in the public interest considering the burden that could result if they were imposed on the facility. ENVIRONMENTAL. CONSIDERATION The staff has determined that the granting of relief from specific ASME Section XI Code requirements does not authorize a change in effluent types. or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this deterraination, the staff i has further concluded that this is an action which is irnsignificant from the i standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4)., that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection wittn the granting of this relief. ACKNOW1.EDGEMENT G. Johnson prepared this evaluation. i Attachmer,t: Technical Evaluation Report, dated September 30.-1982, - t by Science Applications, Inc. Date: June 28, 1983 i i
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7590-01L 4' i ]*, j UNITED STATES NUCLEAR REGULATORY COMMISSION GPU NUCLEAR CORPORATION AND JERSEY CENTRAL POWER AND LIGHT COMPANY DOCKET NO. 50-219 ~ OYSTER CREEK NUCLEAR GENERATING STATION NOTICE OF GRANTING OF RELIEF FROM ASME CODE SECTION XI INSERVICE INSPEC g;N REOUIREMENTS The U. S. Nuclear Regulatory Comission (the Comission) has granted relief from certain requirements of the ASME Code, Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components," to the GPU Nuclear Corporation (GPU) and Jersey Central, Power and Light Company (the lice'nsees). The relief relates to the inservice inspection' program for the 09 ster Creek ' i Nuclear Generating Station (the facility) located.in Ocean County, New Jersey. The ASME Code requirements are incorporated by reference into the Comission's rules and regulations in 10 CFR Section 50.55a(g). The relief is effective as of its date of issuance. c l ( The relief permits GPU to perform inservice inspection examination in 1 a manner or on a schedule different from that prescribed in Section XI of s the ASME Boiler and Pressure Yessel Code and applicable' Addenda, as required by 10 CFR 50, because of inaccessibility, configuration of components, radiation level, or other valid reasons.- The request for relief-complies with the standards.and requirements of the' Atomic Energy Act of 1954, as amended (the. Act), $nd the Comission's rules and regulations. The Comission has made appropriate findings as required by the Act.and the Comission's rules and regulations ~in 10 CFR Chapter I,:which are set forth in the letter granting relief, t T
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The Commission has determined' that the granting of relief'will not result in any significant ' environmental impact and that pursuant to [ 10 CFR 551.5(d)(4) an environmental impact statement or negative declara. tion and environmental impact appraisal' need not be prepared in connection ~ ' with issuance of the relief. - For further details with respect-to. this action, see (1) the applications for relief dated June 15 and Novemoer 15,1978, June 8, September 6, and December 11,1979, June 5 and August 5,1981, May 2 August 2, August 26, and September 14, 1982, (2) the Commission's letter dated June 28,1983, and (3) the Commission's related Safety Evaluation. All of these items are available for public inspection at the Commission's-Public Document Room,1717 H Street, N. W., Washington, D. C. and at the Local Public Document Room,101 Washington Street, Tomms-River, New' Jersey 08753. A single copy of items (2) and-(3).may be obtained upon request I i addressed to the U. S. Nuclear Regulatory Commission, Washington, D. C.. 20555, Attention: Director, Division-of Licensing.. I Dated at Bethesda, Maryland, this 28 day of June 1983. FOR THE NUCLEAR REGULATORY COMMISSION e Dennis M. Crutchfiel, Chief Operating Reactors _ Branch #5 Division of' Licensing. ( 0 h + b a a w r--,-,
4 . SAI Report No. '186-028-34 TECHNICAL EVALUATION REPORT ' OYSTER CREEK NUCLEAR' GENERATING STATION INSERVICE INSPECTION' PROGRAM I i l Submitted to: U.S. Nuclear Regulatory Comission Contract No, 03-82-096 T Science Applications,:Inc.- McLean, Virginia 22102. i September 30, 1982 1 [kh FA Af .M I Science Acces)ecatierrJ,Inc. -h' ~~O -,-e -wi m ---,,-,,e-,- 3 ,-e-- w.,,-4-... 1 y e,
t CONTENTS INTRODUCTION,............................... 1 i .I. CLASS 1 COMPONENTS....................._....... 4- .4 A. Reactor Yessel 1. Request for Relief. Appendix 3A-(Notes 1 and 2), Reactor Vessel Pressure Retaining Welds, Categories 4 B-A and B-8, Items Bl.1 and 81.2..............-.. 4-2. Primary Nozzle-to-Vessel Welds and Inside Radius Sections, Category B-0, Item Bl.4 B 3. Relief Request, Inaccessible Nozzle-to-Safe End Wel ds, Category B-F, Item Bl.6................ 10 4 Vessel and Closure Head Cladding, Category B-I-1, Items B1.13 and 81.14 12 5. Control Rod Drive. Housings, Category B-0. Item Bl.18...... 14 B. Pressurizer (Not applicable to BWRs) Heat Exchangers and Steam Generators (No relief requests). C. D. Piping Pressure Boundary.........,............ 15 1. Relief Request R8, Inaccessible Welds in, Pipe, Dissimilar Metal Welds, Category B-F, Ite.n 84.6........ 15 t 2. Relief Request R8, Inaccessible Pipe Welds in - 17. Penetrations, Category B.J. Item B4.5 3. Relief Request RB, Inaccessible Welds in Piping (But Not in Penetrations) and Supports, Category B-J, Items B4.5 to B4.8; and Category B-K-1. Item B4.9 19-E. Pump Pressure Boundary (No relief. requests)- F. Valve Pressure Boundary (No relief requests) II. CLASS 2 COMPONENTS ......................... 21 A. Pressure Vessel (No relief requests) B. Piping.............................. 21 1. Relief Request R8, Inaccessible Pipe Supports, Category C-E-1, Item C2.5; and Support Components, Category C-E-2, Item C2.6 121 2. Relief Request RS, Class 2 Static Systems 23 1 Al l J science Appucations.ine. ~
1 III. CLASS 3 COMPONENTS (No relief requests) 25 IV. PRESSURE TESTS.......................... 25 A. General i 1. Updating to Newer Code for Performance of 25 Pressure Tests.......'............... B. Class 1 System Pressure Tests (No relief requests)- C. Class 2 System Pressure Tests (No relief requests) D. Class 3 System Pressure Tests (No relief requests) 27 i V. GENERAL.............................. 27 A. Ultrasonic Examination Techniques 27 1. Updating to Newer Code for UT Examinations........ S. Exempted Components (There are no. unacceptable exemptions) C. Other(Noreliefrequests) 29 REFERENCES............................... \\ l l l t ll l ~ l .l t / ~ i Al Science AppHcatens,Inc. - ii-l
y 4 TECHNICAL EVALUATION REPORT-0YSTER CREEK NUCLEAR GENERATING STATION,, INSERVICE INSPECTION PROGRAM INTRODUCTION The revision to 10 CFR 50.55a, published in February 1976, required that Inservice Inspection (ISI) programs be updated to meet the requirements (to the extent practical) of the Edition and Addenda of Section XI of the - AmerEcan Society of. Mechanical Engin'eers Boiler and Pressure Vessel Code
- incorporated in the Regulation by reference in paragraph (b). This updating of the programs was required to be done every 40 months to reflect the new requirements of the later editions of Section XI.
I As specified in the February 1976 revision, for plants with Operating Licenses issued prior to March 1,1976, the regulations became' effective af,ter September 1,1976, at the start of the next regular. 40-month inspection period. l The initial inservice examinations conducted during the first'40-mdnth period were to comply with the requirements in editions of Section XI and addenda in-effect no more than six months prior to the date of start of facility cornercial operation. 4 The Regulation recognized that the requirements 'of the later editions _ s and addenda of the Section XI might not be practical-to implement at facilities because of limitations of design, geanetry, and materials of construction of' 4 components and systems. It, therefore, pennitted determinations-of impractical examination or testing requirements to be evaluated. Relief from theseirequire-i ments could be granted, provided health and safety of the public were not endan-- gered, giving due consideration of the burden placed on the licensee if the requirements we.re imposed. This report provides the basis for granting or denying the various requests for relief by the licensee, Jersey-Central Power, and Light Company, of the Oyster Creek Nuclear Generating Station.- It deals only with inservice examinations of components and with system pressure 1 tests. Inservice tests of pumps and valves (IST programs) are being evaluated separately.
- Hereinafter referred to as Section XI or Code, s_. _._
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j The revision' to 10 CFR 50.55a, effective November 1,1979, modified the l time interval for updating ISI programs and incorporated by refere.1ce a later l edition and addenda of Section XI. The updating intervals were extendesi from i 40 months to 120 months to be consistent with intervals as defined in Section XI. For plants with Operating Licenses issued prior to March 1,1976, Jhe f provisions of the November 1,1979, revision are effective after September 1, l 1976, at the start of the next one-third of the 120-month interval. Durr-ing the one-third of an interval and throughout the remainder of the interval, inservice examinations shall comply with the latest edition and addenda. of Section XI, incorporated by reference in the Regulation, on the da'ce 1r rmonths prior to the start of that one-third of an interval. For Oyster Creek. ?he ISI program and the relief reauests evaluated in this report cover the current 120-month inspection interval; i.e., from December 8,1979, to December 7,1989. j This crocram is based on the 1974 Edition of Section XI of_ the ASME Bai.Ter &: Pressure vessel Code with Adderida throuch Sumer 1975. The November 1979 revision of the Regulation also provides that ISG programs may meet the requirements of subsequent Code editions and uddennda, incorporated by reference in Paragraph (b) and subject to Comission approval. Portions of such editions or adcenda may be used, provided that all rellated requirements of the respective editions or addenda are met.- Thest inst:ances are addressed on a case-by-case basis in the body of this report. Finally, Section XI of the Code provides for certain components an:dt systems to be exempted from its requirements. In some instances, these: taxemp-tions are not acceptable to NRC, or are only acceptable with restricticms. As appropriate, these instances are also discussed-in this report. References (1) to (19) listed at the end of this report pertain te pre-vious information transmittals on ISI between the licensee and the NRC.. In 1976, the Commission providad general ISI. guidance to all licensees. Shsabmittals - in response to that uidance were made by the licensee on June 15,197Bi.,.0) November 15,1978,(2 June 8, 1979,(3) September 6,1979,(4) Muber l'L 19'9,(5): June 5,1981,59) August 5,1981,(10)May2,1982,(12) and August' 2,1952 IIN Pending the results of this review, interim approval was given for.the licensee's ISI p ram on January 14, 1980.(6) The November 15,1978,(2) and Dece niber 11,
- 1979, letters submitted for approval, proposed changes to Oyster Cren!*'s technical specifications.
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NRC requested additional infonnation from the licensee on October 30, 1980,II) and April 26,1962.(11) The licensee responded to those requests on February 5,1981,(8) and July 6,1982.II3), The licensee's response (8) to the first request consnitted to withdrawing Relief Request R3. The licen. see's responseII3I to the second request for additional information comitted to withdrawing the following relief requests:. (a) Appendix 3A; Notes 5,6and10 (b) Appendix 3B; Note 2 ( - (c) Appendix 3C; Relief Requests R1, R2, R4 and R7. 1 Nonc of the withdrawn relief requests are addressed in this report. Two tele-phone connunications between tne licensee and NRC with SAI(16,18) resulted in l clarifying a number of relief requests by the submittal of a new ISI program l plan on August 26, 1982,I17) and a September 14, 1982 letter.III) The licensee has not formal.ly requested relief from the Category B-L-2 l and B.M-2 requirements to visually examine the internal surfaces of certain Class 1 pumps and valves, proposing instead, to perfonn the examinations when-any of the components are disassembled for maintenance. If, at'the end of the current inspection interval, pumps or valves in some groups have not been dis-assembled and examined, the licensee will need to request relief from the applicable Code requirements. Also, NRC has requested,57) and the licensee has implemented (0'I ) an augmented-weld examination program for certain safety related systems that would otherwise be exempt from Code examination by- - provisions of IWC-1220. Frem the above referenced submittals, a total of seven requests for relief and two requests to update to a newer code were identified as requiring disposi-tion. These requests are evaluated in the following sections of this report. i Al J Science ADplicat60ns.inc. -
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- CLASS 1 COMPONENTS A.
Reactor Vessel 1. Request for Relief, Aeoendix 3A (Notes 1 and 2). Reactor Vessel Pressure Retaining Welds, Categories B-A and B-8, Items Bl.1 and Bl.2 ~' Code Reevirement Category B-A:' Volumetric examination erf the shell longitudi-nel ano circumferential welds (in the core 1 region) may be performed at or near the end of each inspection interval and shall cover at-l least 10% of the length of each longitudina.1 weld,'and 5% of the length of each circumferential weld, with ::dhe minimum length of weld- -l examined equal to one wall thickness. i Category B-B: This category' includes vessel shell welds not L in Category 6-A or B-C and all head welds. The -lumetric exami-1 nations perfomed during each inspection irstery, nali cover at least 10% of the length of each longitudinal' - seld and meri-
- dional head weld..and 5% of the length of el mmferential.
3 shell weld and head weld. 1 4 , Code Relief Recuest Relief-is requested from the Code reawired volumetric-examinations of the following reactor vessesi seam welds: (1) Longitudinal and circumferential tshell welds in the-I core region - Category B-A. (2) Longitudinal and circumferential sthell welds-not in the core region except the upper (i ft. of each of.the three longitudinal welds in the upper shell course - Category B-B, (3) All meridional and circumferentiall welds in the lower [ head - Category B-B. 1 Prooosed Alternative Examination Visual examination of all vessel welds during pressure testing i each refueling outage and hydrostatic testfrng each inspection inter-val will be performed. A l scence Apoticahons,Inc..
Licensee's Basis-for Reevesting Relief i-All reactor vessel welds'with the exceptions noted above-i are inaccessible to any local examinations because of. interference with the biological' shield wall. Minimum pressurization tempera-- ture restrictions, the vessel material surveillance. program. themal transient restrictions, and coolant leakage monitoring are specified in the plant's technical specifications._ These provisions, along with the proposed visual examinations,-assure the integrity of the vessel. Evaluation The SEP Report (15) described the-status of. Dyster Creek's material surveillance program and its pressum-temperature operating limits. The material surveillance program was planned before Appendix H.10 CFR 50 was first issued and, therefore, does not. meet all the Appendix H mquirements. The surveillance-program now consists of only two capsules and neither dEe' planned to be i tested by the licensee at the writing of the SEP report. The report recomends that one of the capsules, pmferably theJ2 ~ capsule, be withdrawn and tested at about 12 to 15' Effective Full l Power Years (EFPY). The report indicates that the licensee's technical specifica-tions contain pressure-temperature operating limits that confom to i Appendix G, 10 CFR 50. Since no material surveillance capsules had' I-been tested before the report was written, the operating limits were based on radiation damage' estimates using Regulatory Guide r., 1.99, Revision 1. ~ j The SEP report concludes that low primary stresses 'in-the vessel beltline region (75% of those allowed oy Section III:of the Code) and acceptable material fracture toughness-properties should provide assurance-that brittle fracture will not occur. The ) report then recomends that a material surveillance capsule be withdrawn and tested at about 12 EFPY. The pressure-temperature operating limits should be revised to reflect the results'of the : test. Adherance to the Code requirements would necessitate that p 1 ~ the licensee remove portions of the biological. shield and the per-manently installed insulation to perfom the reovired examination 4 from the vessel exterior of the Category B-A welds and B-B welds listed above. Thus, Code. requirements are impractical because.of existing plant design and geometry. I. - - ~ An alternative examination program should be implemented to - i provide infomation on the reactor vessel's integrity _ that_would represent the infomation gained if all Code requirements were met. The following program for examination of accessible welds. should provide that infomation: (a) The longitudinal welds of the upper shell course should be examined to the extent that the sample size is equal-to that required -for the Category B-A' and'B-B welds for which relief is requested.
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i (b) If the sample size above is calculated to be greater than the total length of accessible longitudinal vessel welds, the closure head meridional and circumferential (Category B-B only) welds should be examined to an ex- . tent greater than normally required by the Code for those welds to make up the difference. (c) The Category B-C (closure head-to-flange and vessel-to-4- flange) welds should not be included in the above samole-size but should be' examined to the full extent of the Code. Additionally, visual examination for gross leakage should. be made as proposed during system pressure tests in the area of the lower head and shield annulus below the' vessel,, Conclusions and Recomendations Based on the above evaluation, it is concluded that for l the welds discussed above, the Code requirements are impractical. It is further concluded that the' alternative examinations dis-cussed above will provide necessary added assurance of structural.- reliability. Therefore, the following is recomended: 4 Reliefshouldbegrantedfromvolumetricexaminationb. the identified welds for the 10-year inspection interval with j the following provisions: l (a) All accessible portions ^of the following reactor i vessel welds should be Code examined during each inspection interval as follows: 1 (1) The longitudinal _ welds of the upper shell' course should be examined to the extent that the sample size is equal to that' required for the Category B-A' and B-B welds for which relief is requested. (2) If the sample size above is calculated to be greater. that the total length of accessible; longitudinal' ves-sel welds,- the closure head meridional and circumfer- - ential (Category B-B only) welds should be examined-to an extent' greater than normally required by the Code for those welds i;o_ make up the difference.. (3) The Category B-C (closure head-to-flange and vessel-to-flange) welds should not be included in,the above sample. size, but should be examined to the fullt-extent of the Code. (b) A material surveillance capsule (pref <rably #2 capsule) should be withdrawn and tested at abcat 12 to 15 EFPY. The plant operating limits should be updated based upon. I those findings.
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.s. ' I i (c) General visual examinations per IWB-1220(c) shou ~id be-made during each system pressure test for evidenc:a of leakage in the areas of the lower head and the sirdeld annulus below the vessel. References 1 References 4, 8,10 and 15. 1 O l 1 4 d l '1 Science Ass:,*sations.inc. 7-emmeemas enum em se
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4-I 4 3 2. Primary Nozzle-to-Vessel Welds and Inside-Radius Sections, Category B-0, item Bl.4 Code Recuirement. The extent of volumetric examination of each nozzleishall-- cover 100% of the volume to be inspected as -shown in Figure IWB-2500-D of the Code. All nozzles shall be-examined during each inspection interval. l l Code Relief Reouest Relief is requested from the volumetric examinations required by the Code on the following nozzles: 1 (a) Recirculation inlet and outlet nozzles: N1A N2A. l N1B N28 N1C N2C N1D
- N20, NIE
- N2E, (b) CRD Hyd'aulic Return Nozzle, N9.
Prooosed Alternative Examination The nozzles are to be visually examined during system pressure tests. Licensee's Basis for Recuesting Relief i There is not sufficient spaceito perform a:UT scan from-4 the reactor vessel wall because of interference from the bio-- 1 shield insulation. Evaluation These nozzles are physically inaccessible, being surrounded by the bioshield insulation, making examination of these nozzles impractical. l Relief is requested for 11 of 24 nozzles on the reactor vessel; the other 13 nozzles are planned to be examined to Code-requirements. All the nozzles are subject to similar environ-j ~ mental conditions. The Code examination of the 13saccessible nozzles and a visual examination in the general area of the lower head during system pressure tests (IWA-5000) should pro-vide adequate informatio'n as to the integrity of the subject nozzles. Science Applications.Inc. Conclusio'ns and Recomendations l Based on the above evaluation, it is concluded that 'for the nozzles discussed above, the Code requirements are im-practical. It is further concluded that the alternative examinations discussed above will provide necessary added s assurance of structural reliability. Therefore, the following is recomended: ) ~' Relief should be grant'ed from the Code required volumetric. examinations of these nozzles.> General vfrsual examinations per IWA-5000 should be required, however, during each system pres-sure test for evidence of leakage in the area of the lower head. Code examinations on the remaining reactor vessel (Category B-0) nozzles-should also be perfonned. s' References References 4, 8, 16, 17 and 18. sam t O I l l l u i-l I science Apphcations.16C. .g. l m-
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- 3., Relief Reouest, Inaccessible Nozzin-to-Safe End Welds, Category B-F. Item Bl.6 i
Code Reouirement The volumetric and surface examinations performed each inspection interval shall cover the circumference of 100% of the welds. -r Code Relief Recuest Relief is' requested from the Code required examinations on the following reactor vessel nozzle welds: t (a) Recirculation nozzle welds: NG-A-26 NG-A-1*% NG-B-25 NG-B-1 V4 NG-C-24 NG-C-1 - NG-0-2 3 ' NG D-1 NG-E-27_ NG-E-1 (b) CRD hydraulic > return nozzle weld: NC-4-1 (c) Liquid poison nozzle welds: i NP-2-1 ( NP-2-2.- l l Procesed Alternative Examina'; ion (1)Visualexaminationduringsystempressuretest,each - refueling outage. -i (2) Visual examination during system hydrostatic test, each inspection interval. Licerisee's Basis for Reoues' ting Relief Oyster. Creek-has nozzle. safe ends fabricated from 300-Series Stainless Steel. The safe ends are furnace sensit.ized. For this reason, during installation.the safe ends-were clad on the I.D. with an overlay _ of 30SL p.terial: and as a result of - surface cracking on the 0.D., were repaired by removing defects-3 and cladding on tne 0.0. with an averlay of 308L material. 1 NUREG-0313 recognizes that clad overlay is an acceptable remedy for stress corrosion cracking. 3ased'on this joint configuration, the-licensee feels that meaningf ul volumetric and surface examination-is impossible. l i AI science Apolications.Inc. l_ a
Evaluation The welds above all have external weld overlays with rougn-surfaces ~. Thh configuration precludes adequate coupling of a-UT probe to the surface and produces a geometry that does not - allow the return of meaningful UT data. Relief..is requested from examining roughly one-half of the Category B-F nozzle welds (13 of 27). The remaining Category B-F welds are scheduled to - be examined, and the examination results from these welds should. provide general informationion the material, condition of all nozzles. Visually examining the general area of the reactori vessel during system pressure tests (IWA-5000) will prov'ide additional information on the integrity of the subject nozzles. Conclusions and Recomendations l J Based on the above evaluation, it is concluded that for the. welds discussed above, the Code requirements are impracti - cal. It is further concluded that the alternative examinations discussed above will provide necessary added assurance.of structural reliability. Therefore, the following is: recomended: i; -Relief shou'id be granted from the Category B-F required examinations on the subject welds. - Relief should be contingent L j upon the Code examination of the remaining Category B-F welds I on reactor vessel nozzles and on the visua1' examination of the general area of the reactor vessel during system pressure testing (IWA-5000). j References References 4, 8, 10, 16, 17, 18, 19 and ' 20'. l -l l i i e l science Apphcations,Inc.,
m. . -.......,..._-7 ] ..l i. 4 Vessel 'and Closure Head Cladding, Category B-!-1, Items BI.?J.3- ~ j and 81.14 Code Reovirement-t 100% of the area of six patches (each 36 sq. in.) of thre e clad of the closure head and accessible portions of the reactor vessel shell shall-be examined each inspection-interval. The ~~ closure head cladding is to be examined (1) by visual and surface -. means, or (2) volumetrically. The vessel cladding is to be i i visually examined, } 3 Code Relief Recuest - I Relief is requested from the Category B-I-1 clad examination l t requirements for the reactor vessel and ci,osure head, ) Proposed Alternt.tive Examination The Category.B-N-1 vessel interior examination assesses the' l general condition of the vessel clad. v Licensee's Basis for Recuesting Relief The Category B-I-1 examination requirements have been= eliminated from more recent editions of the Code because than= cladding is not a pressure retaining component and, thereforte, i not. subject to the ASME Code, u .) l ' Evaluation r The licensee may update. (with NRC. approval) to the 197"J7 Code, Sumer 1978 Addenda, per 10 CFR 50.55a(g)(4)(iv), anat there - e 1 by eliminate the Category B-I-1 examinations.. The licensee !has-chosen, though, to request relief from the 1974 Code provistions and to consider updating to the newer code after the next ran-i l fueling outage. 'Since a portion of the examinations is-rec u ired to be perforced each inspection period, relief would appear to' l be needed. ' The elimination of the B-I-1 requirements from che. j newer code justifies granting relief from the-1974 Code B-IT--I j I l requirements. E Conclusions and P.ecomendations Based en the above evaluation, it-is concluded that f':: ar the eladding discussed above, the. Code requirements, are imprac=5 cal.. It is further concluded that the alternative examination prnoposed 4 will p'rovide necessary added assurance of structural;reliassiility.1 Therefore, the following is recomended. Y M A' I - Sciones Applutaallons.Inc. 12-p s-- ..- ww,.., -io .,_e_ ..--e-s -e = s
Relief should be granted from the Category 8.!-1 recuire. ments to examine the reacter vessel and closure head cladding. l This relief should apply only to that cladding which should have been examined during the first period of the current interval. The licensee should consider updating to the 1977 Code, Sumer 1978 Addenda, to eliminate the need to perfom the examinations during future inspection periods. References References 4,13 and 18. e I i ll AI Scence Appiscations.Inc. 13-M - "6M %.. 9., Q _ h* '. m?' D * --*1----I2-^* J
i: .ap=* .L Control r c DHve Housines, Category E-0. Iteq Bl.ls 4 o Code Recuirement !ne voluT.etric examinations performeo during eacn inspection interval shall include 100t of the wards "n 10; of the peripheral control rod drive housings. The cat.-in6-tions may be performed at or near the end of the insrection interval. Code Relief Leovest Relief is requested from the volumetric examinati:n reovirement of Category B-0 for control rod drive t. sus'ngs. Proposed Mternative Examination Surf ace examinations will be performed in place of volumetric examinations according to the requireo scheduie. Licensee's Basis fc'. Requesting Relief. C:r.p:nent configuration is not conducive t: ve l r. examination as recognized by later editions of the 't Evaluation The 1977 Code, Sumer 1978 Addenda, allows volumetric or surface examination of these welds. On that basis, elief would be justified. However, the examinations may be deferred te the end of the interval and the licensee is considering urcating to the newer code followin The licensee may (with NRC approval)g the next refueling outage. update to the newer code per 1*, CFR 50.55a (g)(4)(iv), thereby eliminating the need for relief. Conclusions and Recom_e_ndations Since relief is not needed until the end of the "-::ecti:n interval, relief from Category B-0 requirements shculo.ct be - granted at this time. Instead of requesting relief, tne Heensee should request to update to the 1977 Code, Summer 1FS Accends, for performing B-0 requirements. References References 4 and 13.
T O B. Pressurizar Not applicable to BWRs, C. Heat Exchangers and Steam Generators No relief requests. D. Piping Pressure Boundary 1. Relief Reouest R8, Inaccessible Welds in Pise. Dissimilar Metal Welds, Category B-F Item 84.6 Code Reovirement The volumetric and surface examinatinens perfonned each inspection interval shall cover the circumfew of 100% of the welds. Code Relief Recuest Relief is requested from the Code rec aired examinations on the following isolation condenser nozzle-t.o-safe end welds: Nozzle NSA to penetration X-5A Nozzle NSB to penetration X 58 Nozzle N3A to penetration X-2A Nozzle N3B to penetration X.2L Procesed Alternative Examination The welds above will be volumetricailly examined from one side only and surface examined according tto Code requirements. licensee's Basis for Recuesting Relief The welds above can only be examined free onc side due to insufficient axial-clearance. Evaluation The licensee will be eximining pipe welds according to Appendix 111 of the Code. Article III-442l0 provides for the angle beam ultrasonic examination (UT).of wlds from one ' side only, using a full V-path. Relief is not :necessary for the welds above. 9 &clence Appheateent.lhC. -.
4 Conclusions and Recorrrnendations Based on the above evaluation, relief is not necessary to perfom the proposed alternative examinations. Therefore, relief should not be granted. I l Refe mnces Refemnces 4 and 8. i S 1 l } i l l l o l I l l // AI sconce Appucations.Inc. ; a ~. w^ ^~ .m.. ~---~e.+.
2. Relief Recuest RS. Inaccessible Piet Welds in Penetrations. Catecory B-J. Item B4.5 Code Reovirement The volumetric examinations perforined during each inspection interval shall cover all of the area of 25% of the circumferential joints including the adjoining 1 ft, sections of longitudinal joints and 25% of the pipe b' ranch connection joints. Code Relief Recuest Relief is requested from the Code required examination of ~ eipe welds in the following 17 containment penetrations: )X-SA,X-5B,X-3A,X-38 p X-6 )X-128,X-70 )X-2A,X-28,X-72 X-7, X-8 )gX-9,X-10 X-4 A, X -4B j)X-61 Procesed Alternative Examination Visual examinations of the areas of these penetrations will be made during hydrostatic tests. Licensee's Basis for Recuesting Relief Each process pipe has a weld inside the penetration assembly-that is inaccessible for any local examinations. The areas of these penetrations are monitored by temperature and radiation monitors that will help detect leaks in these lines. Evaluatien The identified welds are completely inaccessible for volu. metric or surface examination because the welds are located inside a centainment penetration. Each primary containment penetration-assembly, due to its design, leaves one pressure retaining piping weld inaccessible for examination by either surface or volumetric means. The welds can only be examined by inspecting for evidence of leakage during system hydrotests. The initial design of the assemblies did not provide for accessibility for inservice examinations. If it is assumed, though, that the workmanship and quality assurance of the welding as well as the preservice examinations were adequate, then an examination f b 1
c=. of the first pressure boundary weld outside the containneart should reflect service-induced failures for that particular piping section. Thus, the first pressure boundary weld eart-side the containment on each of these process pipes should. be volumetrically examined, where practical, over 1005 of its. length during each inspection interval. Such an examinst$ mn would maintain sample size. Also, the licensee should commeuct Visual examinations at these penetrations as proposed. Conclusions and Recommendations Based on the above evaluation. it is concluded that fuer the welds discussed above, the Code requirements are impras:tical. It is further concluded that the alternative examination cris-cussed above will provide necessary added assurance of structural reliability. Therefore, the following is recommended: Relief should be granted from the volumetric examinat:1on of
- the identified welds with the following provisions:-
(a) The first pressure boundary weld outside the centsin-ment on each of these process pipes should be vollsamet-rica11y examined, where practical, over 100% of Tes length during each inspection interval. (b) The proposed visual examinations should be perfonmed on the containment penetration assemblies when leakage and hydrostatic tests are conducted in accordanca wirts IWB-1220(c). References References 4 and 8. 1 I Al scionee Aass= nations. ine.,
3. Relief Reovest RB, Inaccessible Welds in Pioine (But Not in Penetrations)and Sucoorts Cateoory B-J. Items B4.5 to B4.8; and Category B.K-1. Item B4.9 Code Reevirement Cateoory B-J: The volumetric examinations performed during each inspection interval shail cover all of the area of 25% of ~ the circumferential joints including the adjoining 1 ft. sections of longitudinal joints and 25% of the pipe branch connection joints. ~ Category B.K-1: The volumetric examinations perfomed. during eacn inspection interval are to cover 25% of the integrally i welded supports. l Code Relief Recuest Relief is requested from the Code required examinations of various Category B-J welds (other than those located inside penetrations) and associated supports. Relief for welds located inside penetrations is discussed in I.D.2 of this report. Prooosed Alternative Examination Welds are to be visually examined during hydrostatic testing. Licensee's Basis for Recuesting Relief The subject stelds are inaccessible due to their physical locations either in high radiation areas, very high in a room, adjacent to a wall, or other restrictions without sufficient clearance to perform examinations. Evaluation The number of inaccessible welds needing relief is suffi-ciently small and random, compared with the total number of welds in Category B-J (or in any of the affected systems), that none of these welds needs to be included in the 25% sample to be examined during this inspection interval. For subsequent inspection intervals, the licensee has the option of updating to subsequent Code versions or of staying with the 1974 Edition, throu 10 CFR 50.55a(b)(2)(ii)gh Sunrner 1975 Addenda, pursuant to Updating would allow the licensee to examine the same 25% sample, if the provisions of the Summer 1978 Addenda of the 1977 Edition continue to note (2) of Category B.J in Table IWB-2500-1) prevail (see foot-By adopting -) scente Appletal@n8.IRC. -19' l
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10 CFR 50.55a(b)(2)(ii), the Comission was offering an option whereby " operating facilities with ongoing: inservice inspection programs would have continuity in the extent and frequency of examinations for pipe welds" (see 44 FR 57913). Based on these considerations, relief from these requirements is not required at this time for these welds. It is defer a decision until the next inspection interval, preferable to after the licensee has reevaluated which of the above options he wishes to exercise. Conclusions and Recomendations Based on the above evaluation, it is concluded that for these welds, relief from Code requirements is not needed. Therefore, relief from volumetric examination should not be granted f.or this inspection interval. References References 4 and 8. l E. Pump Pressure Boundary I No relief requests. F. Valve Pressure Boundary No relief requests. .a M v AI science Appheatient.Inc.' 20 9' =.+,g- ^a
!!,' CLASS 2 COMPONENTS A. Pressure Vessels No relief requests. B. Piping 1. Relief Recuest R8. Inaccessible Pipe Supports. Category C-E-1, Item CE.5: and Support Compohents. Category C E-2. Item C2.6 ~~ Code Reovirements Category C-E-1: Integrally welded support attachments are to oe surf ace examined over 100% of each major load bearing element each inspecticn interval. Category C-E-2: All Class 2 support components are to be visually examinea each interval. Code Relief Reevest Relief is requested from the Code required examinations on piping supports and support components for the following con-tainment penetrations: X-66, X-63 X-23 X-20A, X-20B i Preoosed Alternative Examination l None. Licensee's Basis for Recuesting Relief The subject welds are inaccessible due to their physical locations either in high radiation areas, very high in a room, adjae::nt to a wall, or other restrictions without sufficient clearance to perfom examinations. Evaluation The licensee has not provided enough infomation to justify granting relief. The licensee should submit specific relief requests that detail the access restrictions for each support. The infomation provided should include area radiation levels, expected personnel exposures, degree of accessibility for surface and visual examinations for both integral attachments and. support ccinponents. ac6ence Applications.inc. . ~ a e a en emme e _ man.
f Conclusions and Recomendations Relief from Code requirements is not justified and should i I not be granted. The licensee should submit specific relief j requests that detail the access restrictions for each support. The infomation described in the evaluation ebove should be included. References References 4 and 8. e t e l E t 4 s A Science Apphc4tions.inc. 1 ., - - -, - -. - - - ~,. - - - ...,-m.,-....,_ ..,,_,.-e._,_. .,~,,,,,.-..m.,
^-- .8 t' 2. Relief Recuest RS. Class 2 Static Systems t Code Reevirements l I IWC-1220(b): Components in systems or portions of systems, other snan emergency core cooling systems, which do not function during normal reactor operation may be txempted. i j Code Relief Reevest The licensee apparently reovests relief to use the exemption l reouirement for static systems (IWC-1220(a), see below) from the i i 1977 Code, ~5unner 1978 Addenda, rather than the analagous require-j, ment (!WC-1220(b), see above) from the 19.74 Code, Sumner 1975 j Addenda. IWC-1220(a): Components of systems ot portions of systems that curing nemal plant operating conditions are not reovired i to operate or perform a systen function but remain flooded under static conditions at a pressure of at least 80% of the pressure that the component or system, will be subjected to when required I to operate. t Procesed Alternative Examination' Visual examinations are to be perfomed during hydrostatic l ? testing. ] Licensee's Basis for Re:uesting Relief These static syste.ts are constantly subjected to system l pressure and thus have continuous leak checks. Flow induced stress and erosion are not present. Additional testing of i welds in these systems is therefore unnecessary. Evaluation Each edition of the Code has *. different set of Class 2 exemptions (IWC-1220), which treats safety related systems dif-e ferently. Accordingly, to insure adequate surveillance of safety related systems, the licensee should not combine exemp-tiens from both codes. Since the licensee has chosen to base the majority of sxemptions on the IWC-1220 requirements of the 1974 Code, Su we 1975 Addenda, all Class 2 exemptions should be based on that. Code.' Therefore, the Class 2 static system exemption should be according to IWC-1220(b) of the 1974 Code, Si=mer 1975 Addenda, and relief should not be granted. ? sclerect Apphastient. Inc. -
~ Conclusions and Recomendations Based on the above evaluation, relief should not be 9 ranted to exempt Class 2 static systems frcn examination based on IWC.1220(a) of the 1977 Code, Summer 1978 Addenda. The requirements of IWC.1220(b) of the 1974 Code, swuner 1975 Addenda, should be appited if exemptions are desired. References References 4 and 13. O e,
- 6 O
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a III.' Ct. ASS 3 COMPONENTS I No relief requests. !Y. PRESSURE TESTS A. General 1. Updating to Newer Code for Perfomance of Pressure Tests j Code Reovirement l Article IWA-5000 with applicable IWB, IWC and IWC articles applied are to be used for performance of pressure tests. The 1974 Edition of the Code, Sumer 1975 Addenda, was originally l specified as the applicable edition. i Update Recuest Since the original issue of the ISI program, the lican-see has stated that Articles. IWA...IWB, IWC and IWD-5000 from the l 1980 Edition Winter 1981 Adden'da, will be used to conduct pressure tests on Class 1, 2 and 3 systems. i l Proposed Alternative Examination r See above. 1.icensee's Basis for Recuest This action is intended to apply Code articles that were improved over those included in the 1974 Edition through Sumer ./ 1975 Addenda. Evaluation The 1980 Edition of Section XI has been referenced in 10 CFR 50.55a and inservice examinations may meet the requirements of this edition in lieu of those from previous editions with the following provisions: (a) Comission approval is required to update to the more recentedition(pursuantto10CFR50.55a(g)(4)(iv)); (b)Whenapplyingthe1980 Edition,onlytheaddendathrough Winter 1980 Addenda may be used; (c) Any requirement of the more recent edition which is' related to the one(s) under consideration must also be met. e_.. _ _.
7,4 ti ,o e, I l The licens~ee, however, wishes to updarte to the Winter 1981 Addenda, which would require an exemption from regulations. But since the general objective of improving team pressure test pro-j cedures can be accomplished by updating to the referenced Winter 1980 Addenda of the 1980 Edition, this editrian should be used. .I 4 j Conclusions and Recommendations Based on the above evaluation and purssuant to 10 CFR 50.55a (g)(4)(iv), approval should be granted to uppdate to Articles IWA, l IW8, IWC and IWD-5000 of the 1980 Edition, brinter 1980 Addenda, for pressure testing. j References 1 References 4 and 14 i I 8. Class 1 System Pressupe Tests i No relief requests. i i l C. Class 2 System Pressure Tests No relief requests. i D. Class 3 System Pressure Tests ') No relief requests. G 4 8 e 4 AI Sc6enos Appheations.Inc. l 26-1 v + e -s.., ,.,._,,-a.-,.--s.,- .n, .n-, 8--.,,,--,-,.. ..,...r,4 ,,y - -,p , - -, +, ~
i o, ' t. + Y. GENERAL A. Ultrasonic Examination Techniques i i 1. Updating to Newer Code for UT Examinations l Code Reovirement Various articles and subarticles from Sections V and XI as presented in the 1974 Code, Sumer 1975 Addenda. l Update Reouest The liunsee has stated that the following portions of Sections V and XI of the 1977 Code, Summer 1978 Addenda, will l be used in lieu of the same portions from the 1974 Code, ] i Sumer 1975 Addenda. (a)SectionXI: j IWA-2200 IWA-3000 (with applicable IWB, IWC and IWD applied) Appendix III (b)SectionV: l Article 4 l Article 5 Article 6 Article 7 1 l Prooosed Alternative Examination See above. r Licensee's Basis for Reouest to Update This action is intended to apply Code ' articles that either did not exist or were improved over those included or referenced in the 1974 Edition Sumer 1975 Addenda, of Section XI. Evaluation The 1977 Edition of Section XI has been referenced in 10 CFR 50.55a and inservice examinations may meet the require-ments of this edition in lieu of those from previous editions with the following provisions: (a) Comission approval is required to update to the more recentedition(pursuantto10CFR50.55a(g)(4)(iv)); I (b) When applying the 1977 Edition, all of the addenda thrpugh Sumer 1978 Addenda must be used; Science Apphcahens,Inc. - ,,s...w.su, ~.n. ..w.., n., -,.e-e y v. ,p
. a \\ ,:*, 1; s. i I,- (c)Anyrequirementofthemorerecenteditionwhichis related to the one(s) under consideration must also I be met. Updating to the newer Code for the requirements of the above articles would improve the UT examination program at Dyster Creek. All related requirements are included in the update. i Conclusions and Reconrnendations l Based on the above evaluation and pursuant to 10 CFR 50.55a (g)(4)(iv), approval should be granted to update to the 1977 Edition. Sunner 1978 Addenda, for the above subject Code require-ments. References References 4 and 9. l 1 l s i B. Exempted Components l There are no unaccept'able exemptions. C. Other No relief requests. s f = ,W 1 l A 1 seme Awanione.inc., ___,---n.-___,-,_______w_-
. af r -; REFERENCES 1. I.R.Finfrock(JCP&L)toNRC, June 15,1978. .2. I. R. Finfrock (JCP&L) to NRC, November 15,1978. 3. I.R.Finfrock(JCP&L)toNRC, June 8,1979. I 4 I.'R. Finfrock (JCP&L) to NRC, Septainber 6,1979. ~ l S. I. R. Finfrock (JCP&L) to NRC, December 11, 1979. 6. D. L. Ziemann (NRC) to I. R. Finfrock (JCP&L), January 14, 1980. 7. D. M. Crutchfield (HRC) to I. R. Finfrock (JCP&L), October 30,1980. 8. I. R. Finfrock (JCP&L) to D. M. Crutchfield (NRC) February 5,.1981. 9. I. R. Finfrock (JCP&L) to NRC, June 5,1981. 10. J.T. Carroll (JCP&L)toD.M.Crutchfield(HRC), August 5,1981. 11. D. M. Crutchfield (NRC) to P. B. Fiedler (JCP&L) April 26, 1982. I 12. P. 3. Fiedler (JCP&L) to.D. M. Crutchfield (NRC), May 2,1982.- 13. P. B. Fiedler (JCP&L) to D. M. Crutchfield (NRC), undated letter, received July 6, 1982, 14 P. B. Fiedler (JCP&L) to NRC, August 2,1982. + 15. X. G. Hoge, Evaluation of the Integrity of SEP Reactor Vessels, NUREG-0569, Decemoer 1979.
- 16. Telephone conversation: D. Holland (GPU), J. Lombardo (NRC), and
+ R.Yorg(SAI). August 12, 1982. 17. D. Holland (GPU) to D. Outlaw (SAI). August 26, 1982, f r
- 18. Telephone conversation: D. Holland and R. Jeffe (GPU), G. Johnson (tE7E),
and G. Freund and R. Yorg (SAI), September 2,1982. 19. D. Holland (GPU) to R'. Yorg (SAI)~, September 14, 1982. ,, 20. Telephone conversation: D. Holland (GPU) to R. Yorg (SAI), September 20, 1982. Sc6er.a Arm. loc. 29.' t . ~ -}}