IR 05000313/2003002
| ML031330705 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 04/21/2003 |
| From: | Laura Smith NRC/RGN-IV/DRP |
| To: | Anderson C Entergy Operations |
| References | |
| IR-03-002 | |
| Download: ML031330705 (13) | |
Text
Sec>ndsry System Yressure Boundary as an Lxxtension of Lontainment Liner ULD-0-TOP-14, Containment Isolation and Containment Leak Rate Testing (Ref l), states that, AN0 has taken the position that secondary system penetrations are not subject to GDC 57 since the containment barrier integrity is not breached during DBA LOCA conditions. The containment boundary or barrier against fission product leakage to the environment is the inside surfaw of the steam generator tubes the outer surface of the lines emanating from the stearn generator and the outer surface of the steam generator htween the tube sheets This position is based on the concept of treating the secondary system pressure boundary as an extension of the containment h e r Though not well documented prior to the Writing of this ULD, this position has always been the understanding of plant personnel familiar with the original design and licensing basis. Over time, this original design and licensing basis has become obscured by mixing the concepts of the secondary system pressure boundary being an extension of the conraiment liner wiih the concept of a GDC 57 closed systcm. The fact that the two concepts may appear to be equivalent without carefhl consideration has contributed to the cordhion. Before the issuance of the GDCs, nn attempt to distinguish between the concept of extension of containment and reactor building isolation valve was made nor was it necessary. In 1969 (Ref. 21, Bechtel indicated for ANO-1 that in the PSAR, the main steam lines are defined as an extension of containment. About that time an impending ASME code requirement pertaining to piping and valves forming an extension of containment prompted an internal Bechtei memo (Ref 3)
that recommended ail projects provide single isolation valves on main steam and fedwater lines. At that time the ANO-1 design did not include main steam isolation valves With the issuance of ASME Special Rulings #I425 and #1427 regarding piping which forms an extension of the pressure boundary of containment vessels, Bechtef recommended adding main steam isolation valves to the ANO-1 design (Ref. 4). Bechtel later insisted on this addition despite the reluctance of Ap&L, again citing Code Case 1427 and using the phrase extension of containment (Ref S),
This was followed up by Ref, 14 providing additional information advising the addition of main steam isolation valves This recbinmendation was accepted by AP&L in Ref, 15 with the careful insistence that the valves not be called main steam containment isolation valves but be called main steam block valves. That the concept of extension of containment was not distinguished from containment isolation valve was evident in these communications and was unnecessary since the GDCs had not been issued, defining what a containment isolation vdve was and what the design requirements for containment isolation valves were. Once this occurred, the distinction began to appear but only evolved as the full implications of the related GDCs (54, 55, 56 and 57) evolved over the next two or three decade The earliest documented distinction between the two concepts appears to bc WCAP-745 1 (Ref. 61, dated September 1972, which explicitly treated the distinction approximately a year after the issuance of GDC 57. Other examples of documentation of an explicit distinction between the two include NUREG-0830
{Ref. 7j, page 6-1 8 (approving for Callaway the wording of the SNLrPPS FSAR, see pages 6.2.4-5 and 6 2 4-6 and figures 6 2 4-1 pages 1 and 5, and 6.2.4-2); NUREG-OS81 (Ref 8), page 6-1 {endorsing the approvals of hWG-0830 for use at Wolf Creek) and NUREG-0857 (Ref 9) (approving the Palo Verdc FSAR, see page 6.2.4-25). Even with the more explicit wording in the more modern S W P S and Palo Verde FSARs, none of the three SERs referenced above explicitly addressed this distinction; they just gave general approval based on the appiicable FSAR discussion. These facts do not establish a licensing basis for AN0 but do establish the practice and understanding that formed the context &thin which ANO-1 and ANO-2 were licensed
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By early 1972, when Amendment 23 to the AMI-1 FSAR was issued, the response to AEC question 10. l2 indicated that the main steam block valves were designed to Serve as containment isolation valves and that they met GDC57. There was no statement that acknowiedged the applicability of GDC57 to these valves, just that the valves met GDC57. However, later that same year, when Amendment 30 to the ANO-I FSAR was issued, AN0 indicated that they had begun to comprehend the distinction by carefully wording a response to AEC question 5.83 such that there was no acknowledgement that the main feedwater isolation valves were governed by GDC57 but that they were designed to perform the firmtion ofthe reactor building isolation valve in accordance with General Design Criterion 57, 10 CFR 50 Appendix A, after having indicated in the preceding paragraph that the new (at that time) GDC 57 did not apply because ANO-I was designed under older, previously existing requirement When responding to the TMl-2 event in early 1980 in Ref IO, none of the valves associated with the secondary side of the steam generators were listed in the category of autom&xlly actuated valves which provide penetration isolation. For ANO-1, the confitsion appears to have pcaked with the insertion of the GIX:
designations into SAR Table 5-1 in Amendment 1 1 in July 1993 along with a note that read as follows:
These penetrations are associated with the secondary side of the steam generators and are not subject to GDC-57 since the containment barrier integrity is not breached during DBA LOCA CONDITIONS. The containment boundary or barrier against fission produa leakage to the environment is the inside surface of the steam generator tubes the outer surface of the lines emanating from the steam generator and the outer surface of the steam generator between the tube sheet For ANO-1, the following building penetrations have been listed in SAR Table 5-1 as governed by Gm57 at onetime or another. For reference, this table correlates the penetration numbers with the system in which the piping which goes through the penetration is involve Bidg. Penetration #
system Bldg. Penetration #
System PJ, P2 Main Steam P58, P M SG Biowdown P3, P4 Main Feedwater P17, P65 EFW t~ S G S P21, P22, PSS, P63 Service Water No other licensing bask document has indicated that P1, P2, P3, P4, P58, P64, P17 or P65 were governed by GDC57 and only SAR amendments 11 andor 12 and/or 13 included these eight penetrations. Amendment 11 did not include P21, P22, P58 or P64 in this category and Amendment 12 eliminated the above note that Amendment 11 had added and added PSS and P64 to the GE57 category. Amendment 13 added P21 and Y22 to the GDCS7 category and Amendment 14 added back the above note that Amendment 11 had added plus the following sentence at the end of the note Valves associated with these penetrations are not reactor building isolatjon valves Amendment 14 also changed P58 and P64 back to GDC56 and applied the note to PI, P2, P3, P4, P10, P17 and P65, effstively removing the vaives associated with these penetrations front the table Notably, PI0 has never been shown as governed by GDC5 DCD project discrepancy CI-3 (Ref 13) highlighted the lack of documentation of the design basis for the piping penetrations of the reactor building. This lack of documentation was causing confbvion which was exacerbated by personnel with experience at other nuclear power plants that had ownership
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of the Appendix J testing program The lack of design documentation caused design basis questions to be addressed to the Appendix J testing program personnel by default and the answers that were generated were frequently colored by their experience at other sites in lieu of the absent design documentation. Resolution of CI-3 was assigned to design engineering and eventually addressed by the issuance of Engindng Report 93-R-OOO7-01 (R ) in March 1995. One of the many issues that the developers uf this engineering report had to deal with was the confusion regarding the rote of the secondary system in containment isolation that had compounded over the yean This rcport restated the original position that the penetrations associated with the secondary side of the steam generators
. are not subject to GDC 57 since the containment barrier integrity is not breached during DBA LOCA conditions in its configuration note C-57-2 That note dso stated that, Although not directly applicable, the penetration arrangement most closely matches the requirements of GDC 57, which has been conservatively applied This last part of the configuration note explains the appearance of the 57 in the column of the attachments entitled Applicable GDC. for valves for which GDC 57 is not really directly applicable. The implementation of this engineering report included nearly 40 changes to S A R table 5-1 (plus one to SAR $5.2.2.4.1) including several to make the table consistent with configuration note C-57-2 and internally consistent on the issue of penetrations associated with the secondary side of the steam generators Of the nearly 90 changes to SAR table 5-1, all but 25 were specifically exempted fiorn evaluations by Attachment 1 to procedure 1000.131 (Ref 12) including the changes (other than the addition of note 8)
related to penetrations associated with the secondary side of the steam generators The addition of note 3 was treated as a change back to a previous version of the SAR and, therefore, not unreviewed by definition since it had been previously submitted to NRC: as part of the amendment 11 S A R update, The bther changes related to penetrations associated with the secondary side of the steam generators were exempted under Fl (rearranging information to be more easily understood) as an effort to make the individual penetration listings more consistent with the restored note 8 and, therefore, more easily understoo Tn summary, the effort to eliminate inconsistencies and confusion regarding treatment and identification of containment isolation valves is an ongoing one. That effort has been aided greatly by the issuance of the Engineering Report 93-R-0007-01. One of the areas of codhion that the engineering report has been of value toward addressing is that of the application of the concept that the secondary system pressure boundary inside containment i s to be treated as an extension of the containment liner and is itself the containment boundary. The treatment of the changes to the S A R table was consistent with the requirements and guidance that existed at the time and subsequently. Ultimately, the concept of the secondary pressure boundary being an extension of containment boundary was explicitly approved by NRC in the Safety Evaluation Report approving the renewal of the ANO-1 operating license for another 20 years (Ref 16) as follows In drawing LRA-M-237, sheet 1, the redundant isolation valves (SS-1017B. SS-1018B)
for the test connections of the sampling system are not highlighted as being within the scope of license renewal. However, containment isolation provisions require double isolation at the test connections for greater assurance of containment integrity. The staff asked why the second isolation valve on each test connection were not identified as being subject tn an AMR. In its response to the NRC, the applicant states that this penetration is associated with the secondary side of the steam generator, and is not required to meet
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GDC 57 of Appendix A to 10 CFR Part 50. The reactor building boundary or barrier against fission product leakage to the environment is the inside surface of the steam generator tubes, the outer surface of the line emanating from the steam generator, and the outer surface of the steam generator below the lower and above the upper tube shee Valves SS-lQ17B and SS-1018B are not within the scoping of license renewal because they do not meet any of the scoping criteria in 10 CFR 54.4(a).
The staff found the applicants response acceptabl In addition, Amendment 215 to the ANO-1 Technical Specifications included the following words in the basis for Technical Specification 3.6.3. The service water system is the only closed system within the reactor building to which Specification 3.6.3 Condition C applies. Since Specification 3.6.3 is intended to apply to all closed systems, i.e. those with only one reactor building isolation valve, the service water system is the only GDC57 system according to the ANO-1 Technical Specification base References:
1 UL,D-O-TOP-14, Containment lsoiation and Containment Leak Rate Testing 2. Bechtel letter BL-283, dated 2/17/69, subject: Main Steam Line Design Criteria 3. Tntemal Bechtel memo dated 7/11/69, subject: Main Steam Line Isolation Valves 4. Becbtel letter BL468, datcd 8/1/69, subject: Main Steam Containment Isolation Valves S. Bechtel letter BL-602, dated 7 0/27/69, subject: Main Steam Containment lsolation Valves 6. WCAP-7451, Rev. 1, September 1971, Stcam Systems Design Manual 7. NUREG-0830, Safety Evaluation Report related to the operation of Callaway Plant, Unit No.1, Docket S T N 50-483 8. MfREG-0881, Safety Evaluation Report related to the operation of Wolf Creek Generating Station, Unit No. 1, Docket S T N 50-482 9. MJREG-0857, SafHy Evaluation Report related to the opcration of Pa10 Verde Nuclear Generating Station, Units I. 2 and 3 10.OCANO18014 1 1. 93-R-0007-01, Containment Penetration Design Summary 12. 1000.131, 10CFR50.59 Review Program 13. DCD Discrepancy (21-3, Detailed Design Implementation
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14. Bechtel letter BL-750, dated 1/21/70, subject: Main Steam Containment Isolation Valves 15. Letter Harlan Holmes to Burt Lex, dated 3/20/70 16. lCNA010105 Though ILO words in the PSAR expticilly corrobor;lhng this statemeat have been fixmi, 55.6 of the E A R does imply this by &e way that tbc s~csun line and feedwater line buikhg penetrations arc dificrcntiated from othep building pWn That section appears to equate leakage fiom the steam line and feedwater line Wding pene~-&ons with teakage through the co ntaimmt vesscl irsclf by pointing out that other building peoeuatians are grouped and are in penerration areas and leakage fFom these groups of penetrations will be collected and exhausted in a manner that isolates il h m leakage which might occu through the contaimcnt vcsscl itsell-.
This question referred to tbe turbine stop valves as cunaimnt isolation valves even though they couldnt sewe that fuaaion under GDC57 because neither the Line nor the valves were scismic. Therefore. the AEC did nor necessarily imply compliance with the new GDCs when they rcferrcd io mnuinrnent isolation valve ;];,IT 2-L
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Secondary Systes Pressire Boundary as an Extension ofCo&t&nment Liner
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U-0-TOP-14, Containment Isolation and Containment Leak Rate Testing (Ref l), stares that, AN0 has taken the position that secondary system penetrations are not subject to c;DC 57 since the containment barrier integrity is not breached during DBA LOCA conditions. The containment boundary or barrier against fission product leakage to the environment is the inside surface of the steam generator tubes the outer sufice of the lines emanating fiom the steam generator and the outer surface of the
§tam generator between the tube sheet This position is based on the concept of treating the secondary system pressure boundary as an extension of the containment liner. Though not well documented prior to the writing of this ULD, this position has always been the understanding of plant personnel familiar with the original design and licensing basis. Over time, thii original dwign and licensing basis has become obscured by mixing the concepts of the secondary system pressure boundary being an extension of the containment liner with the concept of a GDC 57 closed system. The fact that the two C O R C ~ ~ ~ S may appear to be equivalent without carefir1 consideration has contributed to the cohsion. The NRC did tacitly acknowledge, just prior to granting the ANO-2 operating license, that no post-LOCA isolation of the secondary system flow paths was anticipated by requesting (and getting)
a commitment that ANO-2 raise the steam generator level after a LOCA high enough to wver the steam generator tubes (Re 2). This commitment was to provide a second barrier to containment leakage if the s t a m generator tubes were to leak. If there was already expected to be a second barrier at the valves outside of the containment buiIding there would have been no need for this commitmen.
Tha earliest documented distinction bctween the two concepts appears to be WCAP-7451 (Ref: 3), dated Septmber 1972, which explicitly treated the distinction approximately a year after the issuance of GDC 57. Other examples of documentation of an explicit distinction between the two include NUREG-0830 (Ref 4). page 6-18 (approving for Callaway the wording of the SNWPS FSAR, see pages 6 2.4-5 and 6.2.4-6 a7d figures 6.2.4-1 pages 1 and 5, and 6.2.4-2); NUREG-0881 (Ref. 5), page 6-1 (endorsing the npprods of NUREG-0830 for use at Wolf Creek) and NUREG-0857 (Ref. 6) (appmhg the Pdo Verde FSAR, see page 5.2.4-25) Even with the more explicit wording in the more-modern SNUPPS and Pafo Verde FSARs, none of the three SERs referenced above explicitly addressed this distinction; they just gave gena-al approval b a d on the applicable FSAR discussion. These facts do not establish B licensing basis for AN0 but do establish the practice and understanding that famed the context within which ANO-1 and ANO-2 were iicense For ANO-2, the following building penetrations have all been listed as governed by GDC57 at one time or another. For refkrence, this table correlates the penetration numbers with the system in which the piping which goes through the penetration is involve Btdg. Penetration #
System BIdg. Peaetration #
System 2Pi, 2P2 Main Steam 2P32,ZPW SG Blowdown 2P3,2P4 Main Feedwater 2P35,2P6S Em to SGS 2B7 SG Sample 2P42,2P48 Plant Hating 2P20,2P21, 2P55, L363 Service Water 2P51,2?59 Chilled Water
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For ANO-2, the confusion appears to have begm with the insertion of the old table 3.6-1 of the Tecchnid Specifications into NZTREG-0336 (Ref 7) when the ANO-2 operating license was issued in 1978. The origidly proposed Technical Specifications did not include this table of containment isolation valves. The table in MJEG-0336 was based upon a hand writtem table submitted in Refwnce 8 in response to NRC question 042.32. Besides 2P7, 2P32 and 2P64, that table also listed building penetrations 2P42, 2P48, 2P51 and 2P59 as GDC57. However, Rcfetence 9 describes l d
leak rate tests for both the inside and outside valves for 2P51 and 2P59 and the system drained and vented inside and outside the containment building for all four (2P42, 2P48, 2P51 and 2P59) building penetrations. The tabk also listed both the inside and outside valves for 2P7 as Q3C57 wen though GKS7 can only apply to outside valves. The inside valves for this building penstration are ais0 not operable with a loss of &kite power. Furthemore, the table did not fist any of the valves associated with 2P1, 2P2, 2P3, 294,2P35 M -6.5 evm though their absence fiom this list would exclude them fiom the more restrictive AOT for containment isolation vafves than would apply to six of the eight valves that are otherwise covered by technical specifications. The rest of the valves associated with the semndary system building peoetratons are not othenvise covered at all by technical specifications. The six valves that were listed in the table that are in the secondary system are 2CV-5852-2, 2CV-5859-2, received a CIAS. The ANO-2 FSAR had a table, 6.2-26, which was added in Amendment 23 in 1974, entitled Containment Penetration Barriers that listed all of the secondary system valves and assigned them to GDC 57. The absence ofthe other secondary system valves in ANO-2 FSAR Table 6.2-26 fkom NtTREG-0336 table 3.6-1 is fiuther indication of curifision regarding what constituted a containment isolation valve. There is no apparent logic to explain why the six valves appeared in the Technical Specification table while the other secondary system valves in the ANO-2 PSAR table did not. A reasonable explanation might be that the individual that drafted the Reference 8 table had Iess than 8 thorough understanding of the knction (or even the configuration) of these six valve CV-5850, 2CV-5858, 2CV-1015 and 2CV-1065. Of tho-O
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2CV-5852-2 and 2CV*SS59-2 Similar inconsistencies to those demonsbated above also appeared in the ANO-2 FSAR, Though table 6.2-26 listed building penetrations 21, 2p2, 2P3, -4, 2P7, 2P32, B 3 5, 2P64 and 2P65 as GDCS7, Figure 10 2-3 shows steam traps 2F211 and 22197 with open lineups to the main steam iine upstream of the MSIV This path is open to the piping that pssses through 2P1 and 2P2 and opens through &e steam generators to &e piping W p a s s through 2P3, 2P4, 27, 2 3 2, 2P35, 2P64 and 2P65. Without remotely aperabfe isolation valves on these steam treps, none of these penetrations m l d meet the requirements of GDC57. Furthennore, Section 6.2.4.2 of the ANO-2 FSAR stated A means of leak testing all baniers in fluid systems that stme a containment isolation finction has been provide There were no such meanst provided for the valves in the lines that pass through 31, 292, 2 3, 2P4; 2P32, 2P35, 2P64 and 2 6 5. This was apparent from Figure 10.2-3 In addition, Table 6.2-26 iists building penetrations 2P42,ZP48,2P51 and 2P59 as GDCS?. Huwwer, it is apparent from Figures 3.2-2, sht. 1 and 3.2-4, sht, 1 that these systems are not seismically qualified beyond the inside the containment building isoldon d u e. Since G X S ? relies on the inside pip-hg as 3 barrier to the release of radioactive materials that are generafed by a LOCA, the piping is considered to be a component used to mitigate an accident. As such it is required by lOCFRl00 to be seismically qualified. Therefore, these fwr building penetrations can also not meet the requirements of GDC57 though they are listed lls such. The ANQ-2 FSAR was internally inconsistent. (Note that 2 4 2, 2P48, 2P51 and 2P59 are now recognized as -56 penetrations of contajnment boundary, not GDC57 and nut extensions of containment liner plate.)
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It &odd be pointed out that there are several references in the ANO-2 FSAR to doubie barriers and being single failure proof but the implications of these depend entirely upon whether the word petmion ref= to a penetration of the containment building or a penetration of the containment bundary. Indeed, ANO-2 FSAR section 62-42 refix$ to %ach fluid penetration through the containment lines plate. This expression is more cotlgigtent with the understanding that penetration refers to a p a c t d o n of the containment bouadary. Because of the steam traps shown in ANO-2 FSAR Figure 10.2-3, the claim Gn ANO-2 FSAR section 6.2.4.1) that the containment isolation systems we designed to withstand.%e failure of any single active or passive component without loss of isolating capability and sirniIar asstdons made dsewhcxe can only be true if penetration refm to a penetration of thc containment boundary. It should also be noted that the two times that the NRC asked for a match of p h o n with GDC (PSAR question 5.38 and WAR question 642.37.4) no such match was provided in direct answer to the questio.
By 1980, when responses to the TMI-2 event were being generated, fimk confbsion was we&d when hvo of these six valves were included in Reference 10 in a list of valves to which an SIAS signal would be added. No d e r secondary system valves were included in this tesponse. Again, there is rm apparent logic to explain why these two vdves wede inciudtd white none of the other secondary system valves were. A reasonable explanation might be that these were the only two secundary system valva that received a CIAS, a fad that, in itself is inconsistent with the requirements unless it was just considered to be the canmative thing to do, a practice (going beyond the requirements) that was encouraged as long as it WBS reawnabl By this time, the drawbacks of having detailed component lists in the Technical Specifications began to become obvious enough across the industry that efforts to correct the situation were initiated. The first rcsults of these efforts were Generic M e r 84-13 which faciiitated the remod of lists of snubbers from the Technical Specifications. Lists offire protection barriers and equipmeat subsequently found their exit from the TechnkaI Specificrations and, eventually, Generic Letter 9148 f a c i l W the removal of lists of containment isolation valves fiom the Technical Specifications. For ANO-2, this was accomplished in License Amendment 154 dated Deccmber 22, 1993, which was anticipated to climinate some of the problems with Table 3.6-1 thst had been w n (e.g. License Amendments 108 and 1 12). Izhe SER for Amendment 154 indicated that the list wwld be moved to Procedure 2203.005 (Ref. 27). This list was subsequently reIocated to proceCitlre 1015.034 (Ref 11). The S E R for Amendment 154 dso stated, Overall, these changes will allow licensees to make mrrdom and updates to the list of components for which these TS requirements apply, under the provisions that control changes to plant p r d e s as specified in the AdfiStr0tt;ve Controts Section of the TS UdartumteIy, although the processes were in pLace for these controls, the documentation of the technicat bases that was needed to support these processes WBS mt available. This fact had been identified by ?he DCD project more than a year earlier in discrepancy CI-3 (Ref. 26). This lack of documentation w&5.awing d s i o n which was exacerbated by persomd with experience at otbr nuclear power plants that had ownership of the Appendix J testing progtam. The lack of design documentation cawed dssign basis questions to be addressed to the Appendix J testing program personnel by ddmlt and the answers that w e
generated were fhquetrtiy colored by their experience BL otkr sites in lieu of the absent design documentation. Resohtiun of cf-3 was wigned to design engineering and eventually addressed by the issuance of Engsneering Report 93-R-ooO7-01 (Ref. 13) in March 1995. One of the many issues that the developers of this engineering report had to ddwih was
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the confusion regarding the role of the secondary system in containment isoation that had compounded over the years. This report restated the origid position that the penetrations associated with the sccorrdary side of the steam generators.
~ are not subject to GDC 57 since the containment barrier integrity is not b m M during DBA LOCA wnditiom in its configuration note C-57-2. That note also stated that, Although not directly applicable, the penetration armngement most closely matches the requirements of GM= 57, which has been mnservativeiy appljd This Last part of the configuration note explains the appearance of the 5 T in the column of the s#achments entitled Applicable GDC fix vdves for which GDC 57 is not really directly applicable. The implementation of this engineering report included nearly 100 changes to S A R table 6.2-26 including s e v d to make the table consistent with configuration note C-57-2 and internally consistent on the issue of penetrations associated with &he secondary side uf the steam (genetaton Of the nearly 100 changes to the SAR table, all but 3 wete specifically exempted from evaluations by Attachment 1 to procedure 1000.13 1 (Rd 14) including the cduulges related to pcnetratians associated with the secondary side of the steam generators. The addition of note 9 was exempted under F2 of Attachment 1 to 1OOO.131 (increase in the level of detail without changing the intent) and the other changes related to pmetrations associared with the secondary side of the steam genefators were exempted under F1 (remangins information to be more easily understood) as an effort to make the individual penetration listings more consistent with note 9 md, therefhe, more easily understood. NE1 96-07, Rev. 1 (Ref IS), section 4.1.3 states that IO CFR 50.59 need not be applied to... corrections of hconkstencies within the UFSAR (e.g,, between sections) and this action was consistent with that
~idance. However, the usual practice of producing a discretionary 50.59 ejvalumrtion should have been applied in order to document that undcxmding since the undemanding was based on safety issues that were reasonably likely to be raised in the fbture outside the cuntext of lOCFRsU.5 Later, in 2002, when this was questioned by the resident NRC inspector, a discrdionary evaluation of the changes related to penetrations &associated with the secondary side of the aaun generators was prepared. This evalu&an was prtpared under the 2002 version of 1OCFR50.59 and tfre 2002 Entergy 1OCFRSO.59 procedure &I-101). The evaluation showed that no meviewed safety existed. Since the
~ w p t m c e criteria for dose under the 2002 version of locFRs0.59 is no more than a minimal irtcrease, that is what the evaluation showed. Howevex,the duatian couM also have easily shown that there was m? even a minimal increase if that had been the criteria. Utldef the 50 59 prosfam that was in place when the changes to the S A R table were made, the criteria was a shift fiom one conseqaence category to the next higher or a significant shift within a consequence categor The steam generator tube rupture dose analysis for the ANO-2 S A R prior to steam generator replacement was perfomred in CE calculation 6370-1 1 1240-SQ-TR-001 (Ref. 16) approved on 5/24/77 This calculation was performed wing the base AWO-2 CESEC model and modifying it to simulate a steam generator tube rupture with a concurrent loss of AC power. The changes made to the base ANO-2 CESEC model are listed on pages 12 through 16 of the calculation. An explicit assumption that the steam generator sample lines are open is made to justify the timing assumed for detection of the tube rupture and the subsequent operator actions to lower RCS pressure to a point that termhates the tube leak at % how. lW! of the Xe in the steam generators is assumed to be released tu the atmosphere during this time. In order to cool down to shutdown cooling, the intact steam generator has prmsure lowefed and a pre-existing 100 gpd tube leak is assumed to continue in that steam generator until shutdown cooling conditions are reached at 3.03 There is no Ghange listed in the CESEC model
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that would simulate the c10sh.g of the steam generator sample lines at any time dutiIlg t i i s 3.03 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> This couid only be the result of one thing since all do=
are calculated on the basis of radioactive mllterials being "uansported to the site environs by steam refeased during the transient" (see page 2 of the calculation). Tbat is that releases virt the sample he5 are Ibegiigible within thc context of the calculation. The results of the stearn gemtor tube rupkrre dose anafysis would, -&re, be unafccted by whether t.h sample lines were isoiated at any time during the transient This not only establishes tbat there is not wen a minimal increase in oRsh accident dose but, in conjudon with additional meLtefiaI in the belated 50.59 evaluation mentioned above, establislm a basis for conchding that the sample line valves are not equipment important to safkty with respect to their close fimxiofi A PIP to make procedure 1015.034 consistent with the engineering report was also issued but the c o d m o n checklist &led to i d e e 8 need to change procedure 2305.005 (Ref 12) or the IST cuntdhg document (KES-18) (Rd 17) or th Appendix J testing controlling document (HES-02)
@E 18)- Engineering Standads were not on the coILfig(LTEtfion checklist at the time. The-succts8 of the engineering report in helping with the confusion was almost immediately evident with its use to justify the adminisbative dosure of CR-2-95-0146 (Ref. 19) and CR-2-95-0151 (Ref 20) as doarnmtcd in ANo-95-2-00I 14 w. 21). However, the incomplete configuration checklist led to the cattinuation of some inconsistencies in the treatment of the stearn generator blowdown valves and the steam generator sample whes in 2305.05, Supplement 1, and MeS-18 and its successor document, CEP-1ST-l w. 22).
INPO SEN 97 (Rd. 23) addressed an SGTR event at Palo Vexde Unit 2 in which the functional recovery procedure WBS cntmd because the tube rupture went undiagnosed, due in part to the steam generator blowdown flow isolation on low pressurizer pressure. This lead M a reevstuation of the commitment previously made in I980 to add sn STAS to 2CV-5852-2 and 2CV-5859-2. During the reevaluation which led to LCP 936026 (Ref 24) and the ultimate removal of the SfAS f b m these valves, the engines developkg the LCP contacted the AN0 S s f i analysis group about Ihe acceptability of removing the SLAS. The engineer was told that no credit was taken mywhere for the SIAS other than to meet the 1980 commitment and that any isolation requirementrr for t h e two valves, ;f rury d, would surely d l l be coveted by the remaining CIAS. This conversation was docrzmented inthe LCP by the statement that, "Discqssions with the safety analysis group in Design Engineering indi@
that only the ClAS signal iti credited fix the isolation ftnction" In fact, tme issue of d e h r there was any containment isolation f h c t b n at all was nat addwed in the c o n v ~ m since tkguestion regarding the SIAS did not require if to bc addressed. This OcCwTBd prior to the issuance of the engineering report and might have bcen more accurately worded had the engineering report been available at the dm The wording in LB 93-6026 is in a onetime use document not subject to revision. Supplement I to 2305.005 and page 138 ofthe ANO-2 Appendix to CW-ET-1 are revisable and have been subsequently revised to remove implications that dtber the blowdown or SG sample valves are containment isolation valves or have a containmeat isolation hction. These are the types of inconsisWxim that the engineering report was intended to eliminate. Additional inca&stencies are being identified md addressed under CR-ANO-2-2002-02053 w. 28).
Some s u m s in the mea of clearing up confusion in this area is evident in the design spcxification fbr the replacement steam generator. ANO-M-2557 (Ref 25), Section 304.8.8, states:
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Add bo section 4.5; In summary, the effort to eii-e inconsistencies and codiuion regarding treatment and identification of containment idation valves is an ongoing one That &ort has been aided greatly by the issuance of
&e Engineering Report 93-RaoO7-01. One of the arcas of confbion that the engimaing repor& has b e n of value toward addrcssii is that of the application of the concept that the secondary system pressure boundary inside conthem is to be treated ~ 1 9 rn extension of the mtahment finer and is itself the mntainment boundary. The treatment of the changes to the S A R table was consistent with the requirements and guidance that Cxisted at the time and subsqvently. However, the decision to not prepare an evduation as part of the lOCI.RSO.SS process to document the disposition of the pomthl srtfety issue was neither prudent nor consistent with mnnal practice. Had one been prepared, its
.
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existence could have prevented even firthex coafirsion ia this area in which so much effort has been invested to eliminate confUsio References:
2. 2cAN067837 3. WCAP-7451, Rev. 1, September 1971, Steam Systems Design M i w d 4. NuREG-0830, safety Evaiuation Report &ted to the operation of Callaway Plant, Unit No.1, Docket STN 50483 5. NUREG-o%Il, Safety Evaluation Report retated to the opersrtion of Wolf Crcek G e a d n g Stdon, Unit No.& Docket STN 50-482 6. NuREG4857, Safety Evaluation Report relsted to the operation of Palo Verde Nuclear Gmmting Swim, Units 1,2 and 3 7. NuREG-0336, Arkansag Nuclear One, Unit 2 Technical Spifieations, Appendix A to License No, NPP-6, July 18, 1978 8. 2CANO17708 IO. OCANO18014 11. procedure 1015.034, Containment Pen~alion Administrative Control 12. p d
w
2305.005, Valve Stroke and Position Verification I3.93-R10007-01, Containment Penetration Design Summar-14. 1OOO. 13 1, lWFR5O.S9 Review Pragran~
15. NEINW7, Rev. I, Guidelines fur 10 CPR 50.59 Implematation 16.6370-1 11240-SQ-TR-001, Steam Generator Tube Rupa;lre With Concurrent Loss Gf AC 17. XES-18, ANO-2 E T F r o m BUCS U O c ~ m ~ n t 18. RZ3s-02, Containment Leak Rate T & q pfogram
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20. CR-2-95-0151, invoiving EFW steam supply h e drain lines 21. ANO-95-2-00114 23. INPO SEN 97, S m
Generator Tube Rupture 24. LCP 93-6026, M d Of S W 2 S w
to 2CV-5852-2 & 2CV-5859-2 25. ANO-M-2557, Replace SG Bid Specificstion 26. DCD Discaepancy Cl-3, Detailed I)esign lmplemenration r
27. procedure 2203.005, Loss o C. "
Xrrtegitty