ML20126D340
| ML20126D340 | |
| Person / Time | |
|---|---|
| Issue date: | 01/24/1980 |
| From: | Gower G, William Ward, Wilbur H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| Shared Package | |
| ML20126D330 | List: |
| References | |
| REF-QA-99900400 79-HQ-003, 79-HQ-3, NUDOCS 8004280149 | |
| Download: ML20126D340 (100) | |
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REPORT OF INVESTIGATION TITLE: Babcock & Wilcox NPGD/Possible Violation of 10 CFR Part 21 CASE NUMBER: 79-HQ-003 SUPPLEMENTAL: Vendor Number 99900400 PERIOD OF INVESTIGATION: October 25, 1979 - November 26, 1979 STATUS OF INVESTIGATION: PENDING
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REPORTING INVESTIGATOR: //7N'] / ((4/A'/ 'dNN 2 41939 fard, S'eni'or Invbstigator Wlliam Executive J.O)'ffice for Operations Support, IE:HQ PARTICIPATING PERSONNEL: [ea n r 6 ( ", (/ -
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/H. A. Wilber, Eehior Reactor I~nspection Specialist Division of Reactor Operations Inspection, IE:HQ REPORT APPROVED BY: M67f4 M& M 6eoFge C. @6wer, Acting Executive Officer for Operations Support, IE:HQ e
of 800.4280149
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SUMMARY
l An investigation was condteted congerning B&W Nuclear Power Generation Division (NPGD) handling of two technical issues during the period November, 1978 - January,1973. The two issues as described in B&W NPGD memoranda provided NRC by the Kemeny Commission, were whether a proper analysis had been done for small break LOCAs with Reactor Coolant (RC) Pumps powered and whether a proper analysis had been done for a 10-foot steam generator (SG) setpoint level. The investigation was also to determine whether B&W NPGD's handling of this information was in compliance with the reporting requirements of 10 CFR Part 21.
Investigation at B&W NPGD, including interviews of the authors of many of the documents provided by the Kemeny Commission, revealed that B&W's concerns about these two issues were occasioned by a request from the Toledo Edison Company (TECO) to reduce the SG level from 10-feet to 3-feet in November, 1978. As a result of that request, A B&W engineer noted that the original analyses had been calculated at a 32-foot level; he could find no formal documentation for the 10-foot level although he was aware that B&W's engineering judgment was that the 10-fcot level was bounded by the original analysis.
Other employees subsequently noted that RC pumps powered during small breaks was also unanalyzed. Various employees reported this in internal documents, urged that the formal analyses be done, and suggested that neither NRC nor TEC0 be informed of the lack of analyses. When interviewed, all these employees averred that they did not feel that either issue represented a safety hazard and that the recommendation to withhold this information from the NRC & TECO was an attempt to protect B&W from an NRC overreaction to what they perceived as a technicality. All employees interviewed stated, (two of them under oath) that they felt that the information was not of safety significance and was not reportable in accordance with the existing B&W NPGD Part 21 reporting procedures.
(B&W Part 21 procedures revised in November, 1979 have been expanded to include 3 matters which have " safety implications".) Two employees expressed their belief that TECO was aware of the lack of the formal analyses. The investiga-tion continues in a Pending Status.
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DETAILS PREDICATION AND BACKGROUND On September 10, 1979, the NRC/TMI Special Inquiry Group forwarded a number of copies of documents to the NRC Executive Director for Operations that had been subpoenaed by the President's Commission on the Accident at Three Mile Island (hereinafter referred to as the Kemeny Commission), and inquired as to what actions the NRC had taken from the standpoint of compliance with 10 CFR Part
- 21. These documents and the request were subsequently referred to the Office of Inspection and Enforcement (OIE) for appropriate action. Upon receipt in OIE, the Division of Reactor Operations Inspection (DROI) began a technical review to determine the significance of the subpoenaed documents and to ascertain whether an investigation was warranted. On October 24, 1979, the Assistant Director for Technical Programs, DROI, Mr. Edward L. Jordan, provided the results of this analysis to the Executive Officer for Operations Support and asked that a Part 21 investigation be conducted by the IE:HQ Investigative Staff in view of its apparent interrelationship with similar Part 21 investiga-tions involving the B&W Nuclear Power Generation Division (NPGD) that had been previously conducted by the Staff.
The investigation was initiated on October 25, 1979 with a view to determining the circumstances surrounding B&W NPGD's handling of two issues. The first was whether B&W had properly analyzed the consequences of a small break Loss of Coolant Accident (LOCA) with the Reactor Coolant Pumps (RC pumps) running, and whether their handling of the issue indicated non compliance with the reporting requirements of Part 21. The second issue was whether B&W performed ;
an analysis of a 10-foot Steam Generator (SG) setpoint and whether the handling !
of this related situation represented a Part 21 violation. j i
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INVESTIGATION AT LYNCHBURG, VIRGINIA
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INTERVIEW OF DAVID NMN MARS Mars, a B&W Licensing Engineer, when interviewed November 6,1979 at B&W NPGD, l
Lynchburg provided the following information in substance: B&W NPGD has a set !
of written procedures which implement the provisions of 10 CFR 21. These procedures require, inter alia, that any B&W employee who identifies a substan-tial safety hazard, significant deficiency, or miscellaneous reportable item must file a B&W form called a Preliminary Safety Concern (PSC). Mars described the filing and review process in the following manner. i The PSC must be co-signed by the employee's manager who r6 views it in terms of completeness and accuracy. (Mars was uncertain whether a manager had the-authority to shortstop the PSC). The PSC is then sent to Mars' unit, Licensing, where it is logged and assigned to a licensing engineer. Copies are then distributed to twelve different units within B&W such as ECCS, Nuclear Service, l Project Management, etc. The PSC is then evaluated by the appropriate' technical people within B&W. There is no established time limit for.such evaluation.
An evaluation report is prepared and sent to the same distribution as the PSC -
had been sent originally. It is reviewed by the recipients who must meet an informal deadline of one week in providing comments to Licensing. There is an informal requirement for all such comments to be addressed.
The revised evaluation report is then sent to the Managers of Quality Assurance and Plant Integration for concurrence. Again there is an informal one week review period, but an extention can be requested. If they concur, they sign it and send it back to Licensing which in turn advises ~ the Vice President who i is the responsible officer as set forth in Part 21. He is not asked to concur, but he is asked to acknowledge the notification in writing.
The Licensing Manager either advises NRC telephonically or advises the customer who in turn has 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to notify NRC (if the PSC is adjudged to be a safety hazard).
Prior to Three Mile Island, PSCs were submitted on the average of 20 per year.
Only one of five turned out to be reportable. Mars indicated that there were several reasons why a PSC would be cancelled or determined to be not reportable.
For instance, if the evaluation discloses that the item is not safety related, has already been reported, or the NRC is already aware, the concern would not be reported. A PSC can be cancelled if it is duplicated by another PSC, is already known by a customer who' reports it to NRC in accordance with 10 CFR 50.55e, or if the originator changes his mind. In the latter case, the originator must document his actions.
When advised of the nature of this investigation, Mars stated that he was aware of both issues but was of -the opinion that they- had been properly handled -
in accordance with B&W procedures. He added that the issue of the Reactor Cooling pumps has been the subject of a PSC. 'He subsequently provided a copy of the file associated with PSC 79-16 (enclosure (1)) which deals with the issue of the effects of leaving RC pumps running for.a period during small break LOCAs.
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INTERVIEW OF NIRANJAN H. SHAH Shah, a Seniar Engireer when interviewed November 7,1979, at B&W NPGD, Lynchburg, Virginia, provided the following information in substance: In early November, 1978, the Toledo Edison Company (TECO) raised a question as to whether they could establish a 3-foot steam generator (SG) low level setpoint for the Davis-Besse plant. In researching the answer he discovered that the
, analysis supporting the Topical Report provided NRC (BAW-10075A) was calculated on a 32-foot level. He then wrote a memo November 13, 1978 to Eric Swanson, B&W Plant Integration (enclosure 2) informing him of the foregoing and pointing out that although there have been scoping studies done at the 10-foot level that demonstrate its safety, these had not been reported to the NRC.
He continued to research this issue, and on December 13, 1978, wrote a memo to Lucius Cartin, Plant Integration in which he summarized the data base supporting a 10-foot auxiliary feedwater/SG level control. He noted in this memo (enclosure 3) that it was his opinion that the 10-foot level was safe but a minor cladding temperature excursion may occur. Shah explained that the temperature increase that he cited would be caused by a slight degree of core uncovery, but that the temperature increase would be less than 20 degrees. He pointed out that the increase was hypothetical, a product of the model that wa, being used at the time. He added that a more sophisticated model is in ctrrent use and that he was confident that this model would not show such an excursion at the 20-foot level. He stated that he did not see any reason to consider the filirg of a PSC as he did not view this as a safety concern.
Similarly, about tiis same time, Shah became aware that BAW-10075 did not address the situation of RC Pumps running during small break LOCAs and thus was an unanalyzed issue. He related that he had been informed that such an analysis had been done only for large breaks. Shah emphasized, however, that .
he did not feel that the mere fact that an issue was not analyzed was sufficient to warrant submittal of a PSC as it did not appear to him to present a substan- i tial safety hazard. '
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l INTERVIEW OF ERIC W. SWANSON I Swanson, a Senior Supervisory Engineer, was interviewed on November 6 and 8, 1979 at B&W NPGD. On the latter occasion, Swanson provided oral testimony ,
under oath in lieu of a written statement, this accomodation being reached after the Manager of the B&W NPGD Legal Department declined to approve the furnishing of a written statement. Swanson provided the following information in substance during those interviews:
Shortly after receiving the November 13, 1978 memo from Mr. Shah, Swanson sent a memo to W. H. Spangler, Nuclear Service. In this November 15, 1979 memo (enclosure 4), Swanson characterized both B&W and TECO as being in a "' risk' j position" due to the fact that any indication that the 10-foot level was not '
analyzed, may precipitate re-analysis and re-licensing. He added, however, !
that the B&W ECCS Unit felt that the 10-foot level was adequate. The memo further pointed out that the ECCS unit had not done a small break analysis with RC pumps running, and suggested that if such were done, the results could be unfavorable. He further recommended that an analysis be done for the l
10-foot SG level. l l
Swanson could not recall during interview whether he knew at the time that he authored the memo that the 10-foot level was unanalyzed (notwithstanding the clear implication of the memo that such was the case), but he did realize that !
the RC pump issue was not analyzed. He explained that his remark concerning ;
the possibly unfavorable results of such an analysis were based on his belief 1 that the ECCS needeo a high SG level for proper operation, and that the RC l pumps would create a low SG level that may be unacceptable. He was subsequently informed by Bert Dunn, Manager of the ECCS Unit, that his assumption was wrong. Based on this assurance from Dunn, he felt that the RC pump running l issuc did not present a hazard, and thus did not warrant submission of a PSC.
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Swanfon went on to explain that his use of the term " risk" in his memo did not i refer to a hazard, and that he regretted his choice of that word. He felt I that the point that he was attempting to convey was that it would been have awkward or uncomfortable for either TECO or B&W if the NRC were to demand documentation of the safety of the 10-foot SG level or the RC pumps running in that B&W would be forced to develop such documentation on a crash basis.
Swanson averred that he had no reason to believe that either condition had any safety significance and thus he had no reason to take action in accordance with Part 21. As indicated abov'e, Swanson expressed considerable disnay at l having his remarks in an internal B&W memorandum subjected to such intense i scrutiny a year after they were written, and characterized the effect upon him ;
as being " extremely demotivating".
INVESTIGATOR'S NOTE: Prior to questioning Swanson, the reporting investigator identified himself by display of credentials and informed Swar: son that he was conducting an investigation of a matter within thu jurisdiction of l the NRC. Swanson was further advised that although he had ths. right to not i answer questions, knowingly and willfully providing false information could constitute a criminal offense. Swanson signified his understanding of the j foregoing. '
INTERVIEW WITH LUCIUS R. CARTIN CARTIN, a Senior Engineer, was interviewed November 7 and 8, 1979 at B&W NPGD, Lynchburg. On the latter occasion, Cartin provided oral testimony under oath in lieu of a written statement, this accomodation being reached after the Manager of the NPGD Legal Department declined to approve the furnishing of a written statement. Cartin provided the following information in substance during these interviews:
He acknowledged that he had authored a December 19, 1978 memo to Ray Luken (enclosure 5) that described both the 10-foot SG 1evel. and RC pumps powered issues as not being analyzed and suggested that such information be withheld from both the NRC and the customer (TECO). Cartin explained that he felt that neither issue was a substantial safety hazard falling under Part 21/PSC reporting requirements. He was assured by Bert Dunn that RC pumps powered during a small break LOCA would result in less severe effects than would loss of offsite power. Notwithstanding his lack of concern regarding the safety implications of these issues, he was aware that there was a need for further documentation if for no other reason than the NRC may insist upon it. His comments regarding trying to keep this information from NRC were meant in the context that the NRC might demand such an analysis in an unreasonably short time, and that B&W's inability to respond could result in NRC shutting down or derrating plants. Similarly, to notify the customer would be tantamount to notification of the NRC due to the more stringent reporting requirements that apply to licensees. He emphasized that he was not suggesting, nor did he feel, that information of safety significance should be withheld from the NRC.
Cartin stated that the same explanation would pertain to similar references '
made by him in a handwritten memorandum of January 9, 1979 (not enclosed).
Cartin asserted that Mr. Fred Miller of TECO was aware that the 10-foot level was not analyzed. On the other hand, Miller was aware that it was B&W's position that it was bounded by existing analytical assumptions.
I INVESTIGATORS NOTE: Prior to questioning Cartin the reporting investigator l identified himself by display of credentials and informed Cartin that he was {
conducting an investigation of a matter within the jurisdication of the NRC. '
Cartin was further informed that although he had a right to not answer any )
questions, knowingly and willfully providing information that he knew to be false could constitute a criminal offense. Cartin indicated his understanding of the foregoing.
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INTERVIEW 0F RAYMOND C. LUKIN Lukin, Service Manager, was interviewed November 7, 1979 at B&W NPGP, Lynchburg, at which time he provided the following information in substance: He has been involved with activities concerning Davis-Besse since May 1978, and assumed sarvice cognizance over the facility in August 1978. Upon assumption of those duties, he recalled that the 10-foot SG level was ccqsidered to be the authorized level. He was then unaware that 32-feet had been used in the Appendix K analysis. He learned that upon receiving copies of memoranda written by Eric Swanson and Lou Cartin in December 1978. Although he became aware of both the SG level and RC pumps powered issues at this time, neither raised any safety concerns in his mind. He was aware that the former had been the subject of l scoping studies and that the latter had been looked at at some time in the past. He characterized his feelings at the time as being, "a warm glow" concerning the safety of these two issues, a feeling imparted to him primarily by Bert Dunn, the ECCS Manager. Consequently, he saw no need to intiate a PSC regarding either issue. ;
Lukin stated that he agreed with the concerns that Lou Cartin expressed in his December 19, 1978 memo regarding possible NRC action being taken if it became -
known that neither issue had been analyzed in accordance with Appendix K. He felt that NRC, lacking the assurances that he had gotten from Bert Dunn, might overreact to what appeared to him to be a technicality. He reiterated his belief that this did not represent a significant safety issue and thus was not the appropriate topic of a PSC under then existing guidelines. Luken added, however, that if the same facts were to present themselves in today's climate, l i.e., post-TMI, he would certainly submit a PSC. Luken also stated that it i was his belief that TECO was aware that the 10-foot SG 1evel was unanalyzed '
and recalled that a TEC0 representative named Fred Miller was present during some meetings during which this matter was discussed.
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INTERVIEW OF ROBERT C. JONES Jones, a Supervisory Engineer, when interviewed November 7,1979 at B&W NPGD provided the following information in substance: When questioned regarding his December 11, 1978 memo to Lou Cartin (enclosure 6), Jones explained that the issue of whether RC pumps powered during a small break LOCA has been analyzed came up during a conversation, he could not recall any such analysis although he has since been told by Bert Dunn that an analysis had been done. He then wrote the memorandum in question in order to get the issue analyzed. He claimed that he deliberately wrote the memo in a vague and negative fashion in order to assure funding for the analysis. He said that the analysis was ,
subsequently done by Niru Shah and that although it showed some uncovery, the !
fuel cladding was adequately cooled by high velocity steam. He stated that he l did not perceive this as a safety issue and added that he requires some docu-mentation of a hazard before submitting a PSC in order not to waste time. i When shown Cartin's December 19, 1978 memo, Jones said that he agreed with the implications of the document, i.e., that at the time of the memo, both the RC pumps powered and the 10-foot SG level were unresolved issues. He agreed that I the purpose in withholding the information from TECO was to avoid NRC harsssment ;
which would be occasioned by TECO's mandatory reporting of the informatio; to NRC. He felt that B&W was in the process of doing the very analyses that would be requested by NRC without having to adhere to an arbitrary deadline.
Jones emphasized in conclusion that he did not at any time see that either i issue warranted reporting in accordance with Part 21. !
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t 9-INTERVIEW OF BERT M. DUNN Dunn, Manager, ECCS, when interviewed November 6, 1979 at B&W NPGD, at Lynchburg provided the following information in substance: Although a full Appendix K analysis has not been done for a 10-foot SG level, he is of the opinion that the B&W model supports the 10-foot level was well as the 32-foot that was used in the topical. He described the model as being relatively simple with 32-feet being a rather arbitrary point. Nonetheless, his scoping studies suggested that TECO's request for a 3-foot level could not be supported without further analysis. Dunn verified that this issue came to light as a result of TECO's request for a lower than 10-foot level. He further stated that it appeared obvious that RC pumps powered was a less severe condition than loss of offsite power, an opinion reinforced by his study of the problem both prior and subse-quent to the TMI accident. Dunn emphasized that he at no point felt that either issue represented a safety hazard less a substantial safety hazard and for that reason saw no reason to initiate, or have initiated, a PSC.
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INTERVIEW OF EDGAR ALLEN WOMACK Womack, Manager of Plant Design when interviewed at B&W NPGD Lyrchburg on November 8, 1979 provided the following information in substance: He was aware of the memoranda generated by Lou Cartin and others concerning the issues of the 10-foot SG level and RC Pumps powered during small break LOCAs.
He did not feel that either situation warranted the issuance of a PSC based on comments by members of his staff as well as his own technical perceptions of the subject areas. He noted for instance that the key issue regarding the 50 level was not the level per se, but the parameters of temperature and heat transfer. Similarly, the matter of RC pumps powered did not suggest a safety hazard to him even though he had some concern about the resulting void fraction. ,
He added that he also relied on Bert Dunn's judgment that RC pumps powered was not a PSC.
Womack stated that Cartin's remarks concerning keeping this information from the NRC did not represent an attempt to conceal safety information from the NRC. Rather, they were geared to sparing B&W from a possible NRC overreaction to a matter that was essentially a technicality inasmuch as B&W was already confident that no hazard existed and had already initiated steps to do the appropriate analysis.
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INTERVIEW OF HENRY A. BAILEY, JR.
Bailey, a Principal Licensing Engineer when interviewed No'vember 7,1979, at B&W NPGD, Lynchburg, provided the following information in substance: He recalled having received Robert Jones' December 11, 1978 memo regarding the need for a small break analysis with RC pumps powered, but indicated that it did not have much impact upon him at the time. Although he attributed that lack of impact to possibly bad judgment, he averred that he saw no need to take . '
action and did not see that it was of significant safety interest. He stated-that if it was important, he would have expected it to be in the form of a PSC which it was not.
Bailey stated that he had only a vague recollection of the SG 1evel issues -
that were the subject of several memoranda. He claimed that B&W Licensing was apparently only peripherally involved in that issue whereas they had been much more so on the RC pump issue. As an example of the latter, he called attention '
to the PSC which was described by David Mars during his interview.
Bailey commented that he felt that the prospect of a PSC ending.up in the NRC '
Public Document Room has had a chilling effect upon the use of the PSC system. >
He explained that there was a reluctance to raise an issue to the level of a PSC without doing some sort of evaluation first. Bailey added, however, that B&W NPGD has initiated new Part 21 procedures that in effect lower the threshold for PSCs to encompass anything that affects safety. These draft procedures are enclosure (7) to this report. The draft procedures were scheduled to -
become effective on November 20, 1979.
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I STATUS OF INVESTIGATION The investigation will remain in a PENDING status awaiting review of this interim report by NRC management. ,
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ENCLOSURES Item Dissemination
- 2. Shah to Swanson memo Copy all of 11/13/78
- 3. Shah to Cartin memo of Copy all .
12/13/78
- 4. Swanson to Sprangler memo Copy all of 11/15/78
- 5. Cartin to Luken memo of Copy all 12/19/78
- 6. Jones to Cartn memo Copy all of 12/11/78 ,
- 7. B&W Draft Part 21 Procedures s
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friE B BC0CK & WILCOX COMPANY
'OaER GENERATION GROUP hl DISTRIBUTION ro. ,
J. H. Taylor, Manager, Licensing (2817) .os 663.3
" ' . File No.
or Ref. 205 T4.4 PSC 16-79 vej. Date Preliminary Report of Safety Concern PSC 16-79 May 31, 1979 l..........................,........w.......
Distribution -
S. H. Klein .' - a .L /c E. R. Kane B. A. Karrasch J. D. Agar J. P. Jones G. O. Geissler W. A. Cobb H. A. Bailey B. B. Cardwell (D..iMars D. W. Berger E. A. Womack D. H. Roy C. E. Parks J. C. Deddens B. M. Dunn R. E. Kosiba R. C. Jones mU G. M. Olds A. H. Lazar J. McFarland E. G. Ward K. R. Ellison Record Center
- .- L ., G e c r In . cordance with Procedure NPG-1707-01, " Processing of Safety Concerns," I am forwarding herewith a reported concern on a small break LOCA should the RC pumps go off the line by any means such as by operator action or loss of offsite power.
PSC 16-79 has been assigned to this case.
When my staff has completed its evaluation as to whether a
-reportable concern exists, I will communicate their finding's to you. The point of contact within Licensing on this matter is H. A. Bailey, Ext. 2678.
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v i Jilt / fw Attachment / ,
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.) 11. T avi w gid.,
'p swiiP 20TO: (2 77)
BABCOCK & WILCOX
. MAY 2 91979 PRELIMINARY REPORT OF SAFETY CONCERNS /fsc /(-7Q NANACIR, LICEilSlHG. NPGD D stCNIFICANT OtFICl[NCY DAT[
T4.4 3,0m: _,
Charles E. Parks O sussT ANTI AL sAFriv NAna0 FiLr NO. :
CONTRACT No. -
3.c .,n ,0N: Plant Desian D MISC. REPORTA8L( ITEM PAGE i 0F 4* a:= att i0taistf. sv PACE Nuwata. ANY SUPPORTING INFotMAfl0N/00CUMENTS
$$e, t n3. an0 on wnnCn PLANT was IN[ SAFETY CONC [RN 3lTOTouA EnowLt0GE l$ CU$70MER AWAR(T CY[$INO s 0( a : s e 610 '
wNEN 4 NOW Analysis performed by ECCS Analysis Unit in January,1979 for the 205FA ,
standard plant. *l 10 roua rN0wt:0cc is NaC AwAntT arts e NO urn s Now See attachment 2 cir[a AFFEC1[0 CON 1RACTS (CUSTOM (R NAME AND LOCATION)
Possibly All (177, 205 & 145 FA plants)-
015CAlPTica CF SAFETY CONCEAN.10EhflFY AFFECTED COMPON[NT($), $YSTEM(3) CA ACTivlTY/$UPPLIER, AND IMPACT ON
$AF[TY OF Pt,.NT OPERATIONS 3 See attachment 1 O
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1 013Cil8L Cott[CTitt Atil0h COMPLET10/ic 8[ INITIAT!D i
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h I its,0.sisti ..sf ECCS Analysis / Safety Analysis /PS & Controls samaatuut an0 Dalt 3..dt4eP+
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i ATTACHMCUT 1 g In the ECCS Analysis Unit's and Plant Design's progress report for January, ]
1979, the following item was reported as resolution of a concern over the RC pump status during a small LOCA.
Small Break Analysis with No Loss-of-Offsite power - The 0.05 ft2 break was studied on tne 205 f.A. Flants to detemine the impact of keeping the R.C. pumps on. Results show a much more rapid loss of R.C. inventory relative to a case with tripped R.C. pumps. While it has been determined that the liquid inventory situation is worse -
for a pumps running case, hand calculations have been performed which show that, due to the pumps running, a forced flow, steam cooling situation will exist in the core and will result in cladding .
temperatures of less than 670F. Thus, the pumps tripped case remains :
a worse situation for small LOCA. evaluations. This position will be ,
documented during February. '
Examining this case from the standpoint of being able to withstand multiple failures brings about this concern. While the statement above may be true if the RC pumps remain in operation, the case that was run also shows that h the reactor vessel would contain only ^550 ft3 of water in 10 minutes af ter the break should the RC pumps go off line by any means such as by operator i action or loss of offsite power.
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3 Since the RV lower head is E900 ft , it would take several minutes just to fill the head with only1 HPI pump. (RC pressure E1300 psia at 10 min). The core temperature transient would probably be unacceptable. ki."gh hand cales predict a temperature rise of 300-400 F/ min for the hot pin. Assuming a starting cladding temperature of 700F and a 300 F/ min rise, clad temperature would reach 2200 F in 5 minutes. .The lower head cannot be filled in 5 minutes.
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l One solution to this problem is to develop a signal to trip the RC pumps such as a low system pressure signal or some new signal such as a low level signal )
h which currently does not exist.
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In any case, should a trip signal be installed, a great deal of safety and h ECCS analyses would have to be performed or re-examined. On the other hand, if the pumps are not tripped, unacceptable results would probably occur if the RC pumps should go off line. Further study of this situation is warranted. !
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- ATTACHMEt1T 2 BABC0CR & WILCOX COMPAtlY R GEllERAT10tl GROLIP .
O E.R. KANE, LTCENSINC 4 / A B.M. DU:TN, MANAGER. ECCS ANALYSIS (2138) '
File No.
, or Ref.
Date Telephone Conversation with Zoltan Roszteczy on May 15, May 29, 1979 1979 on Stuck Open p0RV Uith Pures Running and No AuxFeed
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l Mr. Rosstoczy phoned on the afternoon of the 14th with a request that B6W supply.information on the expected results of stuck open PORV small This was to break with loss of.no auxiliary feedvater and RC pumps running.
cover the concern that the analyses provided over the weekend which con-sidered the same case with the RC pumps off may not have been the worst '
case. I responded that in my opinion they had studied the worst case but that the scenario of events would be altered by the different pump assump-tion and said that I would consider these in more detail in the af ternoon and call them back with a position. At SPM I was unabic to make contact with them and actually made the call at approximately 10:30AM, May 15th.
As Mr. Rosztoezy was not available, a discussion was held with I described Paul Norian, the analysis as
~
NRC, Bob Jones of our staff also listened in.
follows: A typical small break evaluation of a stuck open PORV without :
auxiliary f eedvater with RC pu=ps running and with one HPI and realistic Scenario of events: The system would decay heat power levels (1.0 ANS).
involve on a homogeneous as opposed to a separated fluid condition and approach high void fractions; at some time between one and two hours it is (75% is the conceivable that the void fraction could be as high as 75%. )
equilibrium void fraction at a decay heat power corresponding to 3000 '
seconds. However, the evolving system can probably not reach this void To fraction by two to three hours as evidenced by the 'n!I-2 transient.
allow for the fact that THI-2 had operating steam generators whereas this event is without operating steam generators I concluded that 75% could be If at that obtained within the RCS somewhere between one to two hours.)
time the RC pu=ps are tripped, the available, 25%, water would fall into !
two locations, approximately 50*: into the RC vessel and approximately 25%
cach to cach secam generator. This would create a solid water icyc1 in
'ths reactor vessel of 7 feet or a core mixture level of approximately 8-1/2 feet. If the RC pumps-did not coastdown instantaneously, I stated that in my opinion the HPI flow occurring during the pump coastdown would be preferentic.11y distributed to the reactor vessel rather than dispersed '
throughout the RC system and that this flou would fill the remaining 3-1/2 feet within the core region. Thus it would be my cxpectation that no the core most uncovery would take place even if the reactor pumps would trip at unfavorabic time. Further, should the HPI flow not fill the reactor vessel, ths cladding temperature hcatup would be minimum and not result in core P damage. The hestup would be limited to between 400 and 500*F and the re-This situation i sulting peak temperature could not be in excess of 1300*F. )
. . p Ww~
BM Dunn to ER Kanc (h5ubj: Telephone Conversation with Zoltan Rozztoezy on Page Two May 15, 1979, on Stuck Open PORV Mith Pumps May 29, 1979 Running and No AuxFeed ld would last for only about five (5) minutes and af ter that time core covery wou As an over-riding concern, I pointed out that there is no again be maintained. intention within the operating guidelines to lcauseInan RC pump trip d transient and that this is true regardless of pump performance variab es.
other words, I restated our position that at least one pump per loop will run I confirmed that my experience with RC pumps running in high until it dies.
void systems has shown no problems with their performance and that our pump experts indicate no concern in pumping a two-phase fluid.
Our phone call ended with Mr. Norian to passIthis haveinformation not, at thison to Mr.
time, hadRoxztoczy and have follevup telephone calls as necessary.
further~ contact on this issue.
EMD/lc cc: R.C. Jones E.A. Womack C.E. Parks-0 e
i
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, ..-. m . . .
g, y, , i .
NE BABCOCK & WILC0X C0biPAt1Y. .,
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POWER GDCRATICil GROUP db1 4O l ? / '. '
- h. A.1:arrasch, Fhnager, Plant ' Integration '
!!. V. leCarli, Fbnager, Duality Assuranco U*~' J.11. Taylor, Fhnager, Licensing (2817) w 3 63,3 File Ho. 205 T4.4 '
C .: s t - or Ref. PSC 16-79 L
, Date '
- Prelir.inary Report of Safety Concern PSC 16-79 Oct. 24, 1979 g.........,...............-~.*....w..'.4 PSC 16-79 presents a concern for a Small Break LOCA combined with a trip of all RCP at some time after about 2 minutes in the accident.
Pursuant to Procedure No. 1707-01, Licensing has completed its evaluation of the subject PSC and concludes that this is not reportable under the requirements of 10 CFR 50.55(c) or 10 CFR 21.
1hc !.!anager, Plcnt Integration, and the blanager, Quality Assurcnce are requested to review the attached report, signify concurrence or non-concurrence, sign, date, and return this sheet to Licensing within one week of the above date; a detailed explanation should accompany any non-concurrence. Should you require additional information, H. A. Bailey (Ext. 2678) is the contact in Licensing.
._ .nu d g vgI N N b .7., A .
y J. H.Vraylor J1rr/fu -
cc: 11. A. Bailey G. O. Geissler Record Center Plant Integrator Fhnager Action Concurrence Non-concurrence 3
Signature Date t
Qua'lity Assurance Yanager Action Concurrence V Non-concurrence $
Signature A/ -
Date
/,//[/g
l H rl . /.> W'
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.m HE BABC0CK & WILC0X COMPANY ,(
\
l OWER GENERAT1011 GROUP f .. tL-s/_
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Wo d. Vrrasch, Manager, Plant ia rati r
j lB.
E. 7 A )DeCarli
,hhnager, Ouality Assun n
,11. Taylor, Fhnager, Licensing (281 _- sos 663.s j
Fi1e llo* 205 T4.4
'Cu t. r Ref. PSC 16-79
' Date lubj
- Preliminary Report of Safety Concern PSC 16-79 Oct. 24, 1979 l ,,.. i..... .. ...., ... ........ .. ... ..= ui ..ir.
PSC 16-79 presents a concem for a Small Break LOCA combined with a trip of all RCP at some time after about 2 minutes in the accident.
Pursuant to Procedure No. 1707-01, Licensing has completed its evaluation of the subject PSC and concludes that this is not reportable under the requirements of 10 CFR 50.55(c) or 10 CFR 21.
The hhnager, Plant Integration, and the Manager, Quality Assurance are requested to review the attached report, signify concurrence or non-concurrence, sign, date, and return this sheet to Licensing within one week of the above date; a detailed explanation should accompany any non-concurrence. Should you require
' additional information,11. A. Bailey (Ext. 2678) is the contact in Licensing.
,/dd , , 4
,med
\/ J. H.VTaylof JllT/fv cc: 11. A. Bailey G. O. Gei'sler s
Record Center 1
1 Plant Intpgf'ator hhnager Action Concurrence , - Non-c urrence , ,
Signature
+tA<f Date lhfhbff 7
u
% j Quality Assurance Manager Action Concurrence Non-concurrence Signature Date
. i
D - . . .
?
Evaluation of Small Break With No Loss-of-Offsite Power Concern This report documents the evaluation of a concern wherein it is postulated that unacceptable results would probably occur if the RC pumps were tripped
. after running for some period of time during a small break LOCA.
Xdentification The affected plants include all with the B&W NSS. These are:
Oconce 1, 2, and 3 Three Mile Island 1, 2 Arkansas Nuclear One - 1 Crystal River 3 Midland 1, 2 Rancho Seco Davis Besse 2, 3 North Anna 3, 4 Bellefonto 1, 2 UNP 1/4 Pebble Springs 1, 2 Erie 1, 2 .
Greenwood 2, 3 Analysis of Occurrence Recent evaluations have examined the response of the primary system during small breaks with the RC pumps oaerative. During the transient with the
- RC pumps operative, the forced circulatio'n of reactor coolant will maintain the core at or near saturation temperatures (no cladding temperature excursion).
Small breaks evolve to high RCS void fractions due to high liquid (low quality fluid) discharge through the break as a nisult of the forced circula-j tion of reactor coolant (Figure 1).
The RCS void fraction will increase in
. excess of 90% in the short tenn. In.the long tenn, the system void fraction would decrease as the.RCS depressurizes HPI increases, and decay heat diminishes.
T,he RCS evolution to a hig'h. void fraction raises the. concern as to'the !
abili.ty of the plant to successfully sustain 'a RC pump trip by any means (i.e., loss-of-offsite power, manual action, etc.) at the worst possible time during the small break transient. That is, if an RC pump trip is' postulated ,
at a time when the. system void fraction is greater than approximately 60-70%
a core heatup would occur because the residual liquid would not be sufficient to keep the core covered.
A cladding temperature excursion would ensue until com cooling is reestablished by the HPI system. {
.y A preliminary estimate of the impact of the pump trip assumption for some of the cases analyzed shows core uncovery times in excess of 500 seconds '
will occur.
Based on previous small break analyses, assuming an adiabatic O ,
heatup of approximately S F/sec during the uncovery period, the expected peak; '
cladding temperatures for a range of 500 seconds core uncovery will exceed LOCA PCT limit.
Table 2 summarizes the core uncovery period for a spectrum of breaks analyzed. )
For continuous pump operation, the core will be covered and the PCT remains near the saturation temperature during the transient. '
Corractive Action 1
A spectrum of analyses has been performed as shown in Table The 1. .
1 R
results from these preliminary analyses indicated the following:
- n. !
Small breaks with continuous RC pump operation can be i mitigated safely, }
- b. r' If an arbitrary RC pump trip at the worst time must be -
assumed, compliance to 10 CFR 50.46 cannot be shown with present plant equipment, realistic operator actions, and a. ,
single failure. j c.
If an early pump trip is utilized, this action must be 3 completed quickly (1-2 minutes after ESFAS actuation).
If a pump trip is not inii:iated within the specified time l frame, the RC oumos should not betsecured. k Under this #
l circumstance, the operator should concentrate on achieving maximum HPI by ' initiating an imediate cooldown and
'depressurization of the primary system.
As a result of the above preliminary analyses, B&W has recomended to the fars Group that the RCP's be tripped.imediately upon receipt of an ESFAS 3tuation caused by low- reactor coolant pressure. l
. . i 1
j u
Reportabili ty ,
.o Dr. Zoltan Rosztoczy called B&W on May 15, 1979, to request some additional small break analyses related to reactor coolant pump operating assumptions.
This request of Dr. Rosztoczy and a repeat of this request on June 8, 1979, is documented in a letter from J. H. Taylor to Dr. R. J. Mattson (NRC) of .
June 8, 1979,
Subject:
NRC Request for Additional Small Break Analyses.
Dr. Rosztoczy was briefed again by B&W on this concern by a telephone call on July 5, 1979. J. H. Taylor explained that B&U believed that un-acceptable results would . occur if the RCP's trip later during the accident for some breaks in the .025 to .2 fta range. B&W requested a meeting with the NRC Staff at this time. July 18, 1979 was proposed by Dr. Rosztoczy for the meeti'ng.
J. H. Taylor called T. F Novak of the NRC on July 10, 1979, to confirm ,
that July 18 would be acceptable as a meeting date. During this conversation, Mr. Taylor reiterated what was told to Dr. Rosztoczy to be certain he under-stood the purpose for the meeting on July 18.
The NRC was completely inform 3d of the results of the above requested analyses in a meeting with B&W on July 18, 1979. The minutes of this meeting have been distributed by memo from H. Bailey to file 20A3.2 on July 23, 1979.
Following this above meeting, the NRC issued IE Bulletin Nos.79-05C and 79-05C on July 26, 1979. This Bulletin specifies both short-term and long-term actions to be taken by Licensees regarding this concern. Further reporting of this concern is therefore not required, since the Commission has been adequately informed.
l l
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___--mm___ . ____ _________._..__-.__._.._____.__._mm.-
Tn'olo 1. Anelp ic Scopa Eith AFU Arnilobim .
' - RC pump trip G RC pump trip 0 .
-I RO purp trip G 90% void with 90% void with .-
l Continuous RC 2 IIPI's & SG BW of both SG's Pucp cperation 90% void with via ADV's
- I Breakt . with no SG EW no SG EW EW via ADV's .'
2 It?I 2 i gipt 3 2 UPI (ft 2 )
0.025 X X .,
X
,0.05 X X X
- X X X X X X
! 0.075 i X 0.10 X X
~
X X 0.2 IAll breako are 1cented et the RC ptrp discharge.
2 With 2 IIPI's availcbic, ?.5% of the total liPI flow is casumed to be lost'out thh break.
i 3 With 111PI available, 50% of the totc1 IIPI flow 10 casuned to be lost out the break for the f first 10 minutes; after ~0 minutes 3G% is casuced to be lost out the break. '
j E ' ! '
6 g
)!i s
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2 e Table 2. . Impact Asse:sment of Break Spectrum With RC Pumo Trip at 90% Void Break Size (ft2 ) Core Uncovery Time (sec) 0.10 550 0.075 625
! 0.05' 575 0.025 0 .
Notes: 1. Two HPIs available during the transient.
- 2. Core uncovery time 'is the time period following pump trip required to fill the inner RV with water to an elevation of 9. ft in the core which is approximately 12. ft when swelled.
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'P.O. Dcx 12C0. Lf r.chburg. Va. 24505 I.
- .* '~ . Telephone:(C04)301. 5111 .;
4une 8, 1979
... . : l- .- . - - -
..- ~ * - -
\...... .
3r, R. J. Mattson, Director ,'
. 1, t< .- O . , ...l n. . . ; .
'* 4
~-
livision of System Safety ' Regulation . :' ', "*[#
/ '.'d -
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Sffice of Nuclear Reactor .' .. ,j, 7 ", .:
J.S. Nuclear Regulatory Comission .
- 1. -
20555 - :.
licshington, D.C. .
I*
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Dear- Dr. Mattson:
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- ~
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@ubject: HRC Request for Additional Small Break Analyses' - . .
[j
! ' On May .15,1979, Dr. Ro'szt6czy called' B&W to r'equest .some additional J smal. 1 break analyses related to reactor coolant pump operating assumptions.The requested j This' request was repeated in a telephone conversation on June . .
.. - 8.
j analyses consisted of the following: . .
Perform an analysis for the worst case break with the RC -
[
2-
- 1) -
- pumps runnino, ARI available, with normal Appendix X -
I assumptions (single'f ailure). If the RC pumps' trip at the 1
worst time, what are the consequences?
7
- 2) Scme as case /s 1) except no AR1 should be utilized.
1 It should be noted that " worst case breaE" and " worst time are undefined, and the analyses sifould demonstrate that all breaks and times are covered. .
It is my impression that Dr. .Rosztcezy is concerned about the possibility l
of high system void fractions evolving in, this scenario and th\ -
i consequences. ' . . l l
lie intend to discuss this work further hith our utility customers on June 13 and will advise you of the outcome of that meeting. '
j truly yours, Ve p.'
/me,, .e -
- ames H. Taylor -
Hanager, Licensing JHT:dsf i*
- cc: Zoltan'Ros:toczy(NRC)
Tom Hovak (NRC) .
R.~B.Borsum(D&W) .4 .
. 'f.
The Del 40C J.WdCOE Company / [$tablhhtd IB67
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'ER GalERAT100 GROUP g g, ggf f
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l FILE. -
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JU!,, 'g g79 , ,. .
yg )
li. A. BAILEY - LICENSING t V
[ 8 4
'N .
sos e n.s "'
I,
' i ALL ,
/ Flie No.
20A3.2 P' OL'E*f
JULY S, 1979' I lm.io..,.....,................,.w.,..,,.
\
. 5-A and teleconJ. Taylor, was held et al.on Thursday, July 5, between Zoltan Rosztoczy of the NRC The call was initiated by BtJI to update Zoltan on B&W's l followup break analyses.' on our June 8,1979 letter to R. Mattson concerning additional small "
t
}.
i showed that breaks in the range of .025 2 to .2 ftJ.
develop large void fractions Tayldr started the :.' tel if RCP's are allowad to continue operation. He went on to explain that S511 .
believes, based on preliminary calculations, that unacceptable results (PCT) .
3 would in the above occur size if therangeRCP'sof trin concern.later during tha accidan.t.for the larger breaks 'l e
?
M A meeting was requested for the week of July 2Srd' to brief the. NR.C Staff -
prior breaks.
sr11 to any further work.or revision by. B&W of the. operator _ guide. lines for ' ). '
.p Zolttn then launched into a discussion with D&W and the Owners Group in their scheduled meeting on July 19 and 20.
of the items the NRC wanted % to d These items are: '
? i
.4 a) 'Small Breaks - a few outstan ding i'tems -
se '
.. ~ ff b')
Core Uncovery Procedure - What indications are available? What would the operator do? -
4 No mechanism for uncovery was specified.
c)
- l. ' g, Other Safety Analyses '- Look at procedures and gdidelines for '
such thingsSafeties, as LOFW, ~$
etc.SSLB, Overpr. essure Transients Stuck Open s
Secondary *-
9.d
- d' Zoltan added that a NUREG Report was scheduled to be out before the end of July,1979, and it was to discuss the.above items (wi.th resolutions if possible) 's beFor, held thisonreason, July 18, he asked .for the meeting (RCP's on with small breaks) ,
to b :'ui 1979. . . '
back to the NRC. B&W agreed to check with t.he Daners Group and report .
HAB:dsf ~
. '. j, sc: J. H. Taylor * '
E. A. Womack R. B. Davis * .
C. E. Parks Mk' '!
G. O. Geissler *.
R. E. Ham -
L. R. Cartin '
H. A. Bailey * -
'6-(
B. M. Dunn - - '* '
r
- Participants in Telecon '
> :e
.. , . .. M '<. ,
. - . o . . n , i u ., unvur -
- p. O. Geissie
- l'..OISTP.IBUTION
. :p...
7 JUL 10179 ,. .3
/ ,, 7 J.. H. TAYLOR - MMAGER, LICENSING (2817)-
y ag .,.
c
- SDS 663 3 /
Fiie 1 or Rev,lo. 20A3.2 C
// -
205 T4.4 PSC 16-79 o
SPALL BREAKS WITH REACTOR COOLANT PUMPS OPEPATING Date, -
MEETING WITH NRC Jay 10,1979 )
j m. u. . . . . . . . . . . . . . . . . , . . . . . . . . u . .. ,.
t
- I t
DISTRIBUTION H. A. Bailey -
R. E. Ham L. R. Cartin D.. W. LaBelle R. B. Davis
[i. Mars -
B. M. Dunn C. E.' Parks '
J I4ECCeiisl5E E. A. Womack '
3 R. C. Jones .
REFERENCE:
Mamo, July H. A. Bailey to File .Same Tubicct c Abova, dated 5, 1979.
As' a followup to the telephone conversation . recorded in the e.bove. re.ferenced . '
m2mo, I called Mr. Tom Novak of the NRC to confirm that July 18, 1979 would , i 1
be an acceptable time to meet with the NRC cnd review - ' ' .' -
th Mr Nov'ck indicated that it would be desirable to hcVe this mseting in the Phillips Buildin
- for Room .P-422. g starting at.. 1:30 RM on July 18. Presently it is scheduled
. ~ -
During this conversction, I also took 'the opportunity to ' reiterate wha't had that certain cases involving break ranges in 2 the .025 to .2 ftbee range can lead a'fter high void fractions develop in the loop.to unacceptable peak cl -
.to !.!r, Novak was to be certain that he understood the purpose for the m and'that Dr. Rosztoczy had passed on this message to him. -
By copy of this memo, Mr. Geissler is requested to arrange for ..
this meeting this week - preferably Friday, y'run for, i j -
- fHT:dsf '
t i
.*, . +. ;
- i
, j e i
. ,si . ;*e1
. W t
GEllERAT1OU GR03P. -
fit'E ,
/.A.CAILEY-LICENSING'(2G78) 11 , sos m.s j f Fil e l'o. I O'.HERS GROUP or Ref. 20A3.2/LS-7 i
Date .
Sl%LL BREAK UITil RCP'S OPERATING NRC MEETING ' JULY 24, 1979 l
ua.,,,,,,.....................w...w.
A meeting was held with the NRC Staff on July 18, 1979, in Bethesda. The .
Owner's Group was also represented by several attendees. The purpose of the m2eting was to report to the NRC Staff on the v.ork done on the NRC request for additional small break analyses with RCP's tripped at the worst time.
.l!
This request was documented in J. H. Taylor's letter to Dr. P.. J. Mattson of ,
June 8, 1979 (attached). Those noted (") on copy distribution were present .
from B&W. The slides used in B&W's presentation are attached. j The G-Node model and the HPI assumption of Toss of all ECC to the broken leg after RCP trip was explained. -
The void fraction y required prior to RCP trip 'to uncover the DB-1 core was discussed. B&W has done no specific calculations for DS-1, but feels it would be higher than the 63'4 required to levar the level to 9 feet of colicpsed liquid in the 177FA lowered loop plants. , ,
The NRC Staff (2oltan Rosztoczy) csked if B&W had taken liquid carryover into '
account. Anseer uas no, but we would expect better cooling if we did. Zoltan then asked that BLW look at low flooding rate FLECHT tests and extrcpolate from that and see if it increases uncovery time.
The ability of the pumps to run during high y was mentioned. B&W cited the previous submittals on this, subject. l l
Dr. B. Sheron of the Staff suggested the window of break sizes might be larger -
l due to the separation of water downstream of. the RC?. He also asked if a i partial loss of RCP's had been looked at. Answer was no. He asked if oump l degradation effects were considered. B&W 'the CRAFT Code.has a degradation modal, but these effects are not significant.
The size of the breaks was noted to be nonmechanistic except for the .025 ft:
(spray line). Zoltan asked about the Code Safety size. B&W - about .02 ft 2, Zoltan then noted that two open code safetics would put the break in the window.
s .,.,,
. i.
O
. l j----
Page 2 -
L July 24, 1979 ile * .
y , . ,
/~
1 The uncertainty of the calculations and how much margin are presentwere Zoltan's chief concerns. Hu pointed out that the 1.2 Atl5 on decay heat was really the caly Appendix K conservatism working here. The 1.2 ANS was shown to shorten core uncovery in one case.from 600 see to 400 sec. ,
i Zoltan pressed B&W for a'reco=and'ation, but was told the owners would make I any reco=2ndation. BDi went on to explain we were looking at tha feasibility of a coincident RCf trip based on low RGS pressure ESFAS and void fraction.
B&W promised the S'.aff an official submittal of the presentation by July 27th.
An SER for the BD1 operating plants has been written, but this new informa- -
tion will be factored into the SER by the Staff. ;
HAB:dsf .
Attach. -
I ,
~
cc: D. H. Roy ,
J. H. Taylor .
E. A. Uomack* .
C. D. P. organ ,
. J. J. Cudlin B. M. Dunn R. C. Jones -
w/o attach.
- L.11. Cartin* -
3 C. E. Parks * ,
G. O. Geissler
- H. V. Bonaca* .
'D. F. Hailman* .
H. A. Haghi* .
. R. E. Ham *
- E. W. Swanson*
4
- Attended Meeting -
e
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} )
j
', ? /. /1,. ,f,A._. . .. a FSS ;l
" i
- nmnoi, c . ;c:
U nTED STATES i
k dp 9 - f'- r;UCLEr.n Rari,TcP.y CCm:s5Icx JUL 271979 b 3
, e( y)
FFICE OF INSPCCTION l;.0 EWO . CEMENT 2.' .pA d W.SHI5GTO:i, D.C. 20555 ;
) ,
i e
July 26,1979 ,
j
.;*...~.
p h.. . j;
,;; .. ., },y () ); IE Bulletin Nos.79-05C & 79-05C
'i UCLEAR %. ItiCIDELTglT W - Td" MILE ISt.A!io - SUFF1.EMbiT y scripticn of Circer.:ttnces: J a
nfenntion has br:ccoa available to the EC, subsequent *to the issusace of .]
1 E Bu11ctins 70-05, 70.05A,70-053, 79-05,79-05A, 79-06A (Revison 1)
~ d 79-053, which requires modificction to the " Action To Be Taken Dy :1 ,
. ?
h.icen:ees" portion cf IE Sulletins79-05A, 79-03A and 79-053, for all.
rcasurized warcr reactors (PURs).
tem 4.c of Bulletin 79 05R required all holders of operating licenses for -
tbccch & Wilecx designM FURS to revise their 00ersting procedures to speciff Int, in the event of high pre.ssure injection (HPI) initiation with reactor .
calcut pu:aps (RCPs) op3 rating, at least uno RCP per leap would re ain operating. -
{sont:ined in Itua 7.c of Bulletin 79-05A (for Westinghouse designed p .
On Iten S.c of Bullatin 79-053 (for Cocustion Enginacring designed plant:). -
Pric/ to the incidtht ut Thrza Mile Island Unit 2 (TMI 2), Uestinghouse cnd -
.its lic.cnrces genercily adoptad the position that the apara:Or thculd promptly Tnis
~
l trip all operating RCPs in the loss of coolant accident.(LOCA) situation. .c' Mestinghouse position, has icd to a series of ribetings betwesn the.fiRC sttff In and .,
Westin; house, as well as with other PWR vencors, to discuss this issue.
addition, nere detailed analyses conce:ning this cr.tter v.are requested by tha 1 ,
Recent preli5inary calculations perfors:d by Babcock & 'n'ilcox, Westing-
. l(RC. .
Cor.bustion Engineering indicate that, for a certain spectrin of
- .' ..g%
~
. house and )
reall breaks in the reactor ccolcnt systa:3, continued operation of the P. cps can increase the mass Icst through the bronk and prolong or aggravata the ur, cover-
"4 j
.9 i
. ing of the reactor core.
n 2. l i !
The darage tr> the ecactor core at 1El 2 foliced tripping of the last operatinD F- a RCP, when two phast fluid was being pr. pad throuch the reactor ecclant systen. ,C.l.
It is our current understr.nding tm.t all three of the nucl' ear stem synom ,.
suppliers for PWRs now agree that an acccptatic nction under LOCA tynstoms :-: .v is to trip all outrating F. cps in adiately, beicre significant voiding in the **
ranctor coolant syst:.1 cecurs,
, '(-
Action To Sc Token Dy Licenscesi C '~ " f.<
In ort!cr to P.11cviate the cencern over delayed tripping of the RCPs after a ;
, LOCA, all holders of operation' licenses for Na fccilities snali tehe the l ~ felle.<in.1 cctiens : f ,
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~ .g TE Duiktin I:os.79-05C L 79-0GC July 25,1979 r-z.
- Pasa 2 of 3 P
.w ..
.f-.,
Short-Tem Actions .
. 1. In the interie., until the design change required by the leng'-w i :g
- action of this Bulletin has bec.n incorporatad, institute the follosing actions at your facilitics: ,
1l
>* 'A. Upon reactor ' trip 'and initiation of HPI caused by ice reactor
- j
- .c. ccolant systen pressure, imediately trip all operating RC?s. -
1 i
1 B. Provide tuo licansed operators in the control roem at all tices 1 during optration to accomplish this action and other irmediate cud folicwup actichs required during such an occurence. For
.; facilitics with dual centrol rc:ms, a total of three licensed .
.' .. - operators in the dual centrol room at cIl tir.us r,2ets tho. require-
,t, trents of this Bulletin.
II2. ~Parfom cnd submit a report of LOCA analyses for your. plants for a
.- range of sc.all brenk sizes r.nd a rcnge of tire lapses between reactor trip end pu.np trip. For each pair of vr. lues of the pcrameters, deter- '
~
mina the poch cindding tr.rpcrature (PCT) which results. The ranga of values for cach parameter mst be wide enough to assure that tha j
c t.nnimum PCT or, if necrooritto, the reninn containino PCTs gt:ater than -
2F.i3 degras F is iticativmu. .
a: .
- 3. Based on the analycos dona under Item 2 c.bove, develop nw guidelines 4v. for operator action, for bai;h LCCli and non-LOCA trensients, that tche into account the impact of RCP trip recuirr.e.nts. For Sabcoch & ,
y;' liilcor, designed reactors, such gui:'alin s should include appropriato ,
requirstents to fill the st2tm sqnerators to a higher level, followinD
?,.
.J RCP trip, to prcccte natural circulation flow. J 4, Revise en rgency procedurcs and train all licensed rcactor operators .
and senior reactor operators bacc4 on the guidlines developed under k Item 3 above. . .
- ..-0 s . . .
.- 5. Provido analyses and develop guidelines and precedures related to in- .
adequate coro ecolina (cs discussed in Section 2.1.9 of INREG-0578, "IlF2 Lessons Learned "f ask Force Stams Rcport and Short-Tent Rce.0:n-
- T.
H
.- mendations") and define the conditions under which a restart of the F RCPs should ta attetpted.
-j{
' Long-Tem A: tion
- y
.2 Propose and submit a desinn uhich ulll assure autr.: tic trisoing of .:
- 1. .j the operating RCPs under all circu: stances in which this action r.ay be sicoded. i
- s .4
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I2 Bulletin tics. 79 05C & 79-05C -
July 26,1979 2
'Page 3 of 3 lj y
i 3-
. Schedule .
..- , , Q.
The schedule for the short-tem actions of this Bulictin is: .
ci 1
Itc= 1: Effective upon receipt of this Bu11ctin, -
9 Ite:n.2: Utthin 30 days of roccipt cf this Bulletin,
,- 1
.Itea 3: Within 30 days of receipt of this Bulletin, .,
. i .
Item 4: Within 45 days of roccipt.of this Bulictin. -
Item 5: October 31,1970 (as noted in Table D-2 of N'JREG-0570.. -
fl i
underItem3). .
t A schedule for the long-tem action rea.uired by this Bu11ctin should ba (
developed and Submitted within 30 &ys of receipt of this Bulletin. .
P.spcrets should be sub:aitted to th2 Dirt: tor of the appropriate f(RC Regional a I - 0,ffico uith copics forwarded to the Director, Office of Insp:ction and Enforen.m n: and the Director, Office of !;ucitar Ret:: tor Regulation, Wa D. C. T.O.:.55. .
] ,
7pproved by G'10 (80372): clearance c::pirer,7/31/C0. A j o bicnkat cictrcnce specificdlly for 9-r.cric pr.nlems.ppmynl ut.s gi'ica unde
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SABC00K & WILCOX COMPANY ER GE" ERA 710N GROUP l
I . t.~. Swanson, Integra:1on i
g g',p ruli . .r" iWwis (213 6)
. - - - - a ses 663.3 FiIe No.
TICD-1 or Ref. NSS-14/~3.4 Date Auxiliary 7eedva:c; Level Centrol Mai-V 13M lni..........,........,...,.
A questics is asked whether a 6.cui'SCIdMEset poin: for the ~ ICD-1
. plant veuld be adegua:e. ~he 2005 Unit has eviewed the f.npact.ptiov) leveren-LOCA_ana2,rg?Q The auxiliary #eedva:er level con:::1 is E@_%cany; fey;7@MeaEs enly. '~he analysis !c: the presen:1y appreved s=all_ creak c pical repe g 3AW-lC073A.,Rev. 1,'Oas based'on a 32 '" Fl ' *"JC-J Subsecuentapcep_nd gdg denU'ti,tIno: repor ed to 'N?d have shewn ths: aji con:re W ade g g c assura ccre_Sait.*y #m- n 7 " j*= -
( d AUT level
- r_ansien:.
Howeyer, arlevel sHutuuseitThii'IG;ff~@u requfri~. add'.iEbi' c_
,sa 4Lgsmt pamm a wn%e av m *-Mwarenhanges-.ceeding 10tC _. ,J. .ap... . _
NES/lc ec: M '
R.C. Jones G.I. Anderson 3.A. Karrasch
- R.C. Luken W.H. Spangle:
I.A. We:aek s
W
G k
W
BABC0CK & WILCOX COMPAllY fERGENERAT!0NGROUP l
L.R. CARTIN, PLANT INTEGRATION h
, /VHS * " ' ' '
N.H. SHAH, ECCS ANALYSIS (2136)
File No.
.t. or Ref.
TECO NSS 14 Date J.
DECEMBER 13, 1978 10 FEET AUXILIARY FEEDWATER LEVEL CONTROL
- m. i.n., ......,..........,............,..#
l The.following small break analysis base exist for TECO plants using auxiliary feedwater level control of less than 32 feet.
In these analyses it was assumed that the loss of offsite power was coincident with the reactor trip thus initiating the RC pump coast-down, main feedwater coastdown and isolation of secondary side of steam generators from the secondary loop.
(1) CFT line break:
AFW 1evel used: 10 feet.
Status of Calc File: A formal calc file number does not exist as the work was never intended for a formal submittal to NRC.
The analysis was done for sensitivity studies for new analy-tical techniques. A loose compilation of calculational input is available.
Status of Design Memo: A Q/A'd memo exist (no design memo num-ber) which describes the analysis and results.
R.J. Salm to H. A. Bailey, "Re-analysis of CFT Line Break,"
September 14, 1976.
Status of Results: Core always remains covered by a mixture thus no cladding temperature excursion occur and core remains in safe condition.
(2) HPI line break:
AFW 1evel used: 10 feet.'
Status of Calc File: Calc 32-4194-00 supports the anal; 's. -
It is fully Q/A'd and released via a DRN.
Status of Design Memo: A Q/A'd memo exist, without a design memo number, which describes the analysis and results.
W.L. Bloomfield to D.B. Tulodieski, "HPI Line Break," T3.4, NSS-14,25,26, January 4, 1977.
Status of Results: With an operator action within 30 minutes, the analysis shows compliance with the Acceptance Criteria of 10 CFR 50.46. The core is always covered by a mixture during the transient.
5% {T)
NH Shah to LR Cartin Page Two
Subject:
10 Feet. Auxiliary Feedwater Level Control Dec. 13, 1978 (3) 0.5 ft: at pump discharge:
AFW level used: 20 feet.
Status of Calc File: Calc File 32-4518-00 is Q/A'd and released.
Status of Design Memo: A Q/A'd memo, without design number, exist which describes the analysis and the results.
M. DiQuarto to D.R. Tulodeski, " Davis-Besse 1, 0.5 ft2 Small Break Analysis," T3.4, February 10, 1978.
Status of Results: CRAFT and FOAM code analysis were needed to show that the core remained covered by a mixture thus maintaining core safety.
Note: It is my opinion that with a 10 feet AFW level control, the core safety will comply with the Acceptance Criteria of 10 CFR 50.46 but a minor cladding temperature excursion may occur.
If we use the presently approved small break analysis model modifications (i.e., two node inner vessel and appropriate phase separation multipliers for all reactor vessel nodes),
I feel that the core will remain covered by a mixture and no cladding temperature excursion will occur.
WSH/lc cc: B.M. Dunn
. R.C. Jones G.E. Anderson H.A. Bailey E.W. Swanson E.A. Womack s
- _____________________.________f__
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- E. Spangler. Nuclear Service .
- m I. 2 Sva f.cs, Plan: ~ :egra:icn :s es:.
.st. ~i l e he.
T: led -NSS-14 J Dj . Date A=111ary Teedvate: Se:pci :s Neve:b er if 19 ~ 5 j n...,......,...................,,.
Our recen: discussions vi:h Tcledo perso= el_regar, ding .hei:. 0p.14.. o :e_cuci .
L::= s eea:Artentrato:F. Level- seve1= ' fer =ansertettir: d =-d er:,iand .'65*.L's eeeA :25 1:ainzaEla)):pDClev'at.itm:;ausd:m i =PJMr have ledi:n?in 1 passa; NUG.,And ~9.le.nR /-sa in.
"-' *h"...Ro.sige-Zb eguitedeh!^ * *h e-> breAid Sopica,'.. was. based _g;) 32' 14 vet m ei-i m a s c'- r Ie_:o da:.pos.1.?q w
,1e:;ui:2.,,::,2:a.:,al,gis.,,,A*,( ,,p.e. .lf,j essi.sg . Nevartheless,,,a s:ea:J,enera:::, lave'
. va '.ne _has..::: been repor:ed :0 ,:QC3 ,, and. de ICOS " .it..be2 eves ,dn:, a 10 ',i oleve* sc:;:i;:lil' be _adeeut.7,;,. __
r-- - ._
N ah '= _ser.Me_ levea,,,;hgJeeinmem-_'tf and I cf f er :he f =11cving suggestien which you should pursue vid Toledo:
- 1. A1:e: the con::c1 logic of the 57RCS so tha: 1: 112 hypn w- FNM Since a con::ci func ics canno: be readily placed is an ISTAS sys:e=, :he STRCS nus: be =cdified. 1: de presence of an ISTAS signal, de ISTAS se:s a pricrity for :pera:icn over a y STRCS signal and diree:s the STROS
- o provide a high se:poin: level con::el. In de absence of an ISTAS signal, bu: vi:h an STROS genera:ed signal, de 57?.05 cen::ci se:pcin:
is direc:ed := a icv level. A general sche =a:ic is a:: ached; c:he: =ethods of i=plemen:ing are possible, bu: dis purveys de eenecp:.
- 2. ISTAS could also ini:iate auxilia:/ feedva e and isola:e =ain feedva:er.
Further investiga ica needs to be cade as :o de actual seq.:ence of even:s.
believe 1: is new possible fe; :ve condi:1cas :: exis because de ~I:s sys:e=s do no: ini:ia:e A'"4 by ISTAS. '"hese are:
Ch ren: Oesi:n Si:e Cendi:10 , Svs:e=s Secuence Cent:51 Se::ei :
- 1. Offsi:e ?:ver Available ISTAS
. a= ICS 2 ' (.v.ain Teedvs.:er)
- 2. Offsi:e ?:ver Unavailable IS7AS > STRCS 10' (Aux. Teedva:er)
- ! : reasening is ::::e::, :he firs: c: d:.1:n viZ : :.y pr vide a 2' cen:::1 (:f _ain fsecua:er); :: STROS signal v. n ec:ur and d e ICS vi n
- :.. he :de ::ndi:icn vili cr.use :he STT.:S: :espond :: a 1:ss :f
.evel (nes: likely) or :c a less f pump ;:ver. A: any :::e. !?ROS viZ ini:ia:e A'".* and ::::::1: 2e ni;h se:pcin:.
f n
. Cf N l
/ ,/
Swanso: :: Spangle: Page h e Auxiliary Teedva:e: Se:pein.s Nove=ber 13, 1873
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I sugges: tha: ~I o c:sfi:= ca: de abcve sequences are corre : bei= e a decision is =ade to ini:iate AW wi.h ISTAS.
3.M)g,ingQg :::.:ir= ;;.a g J6 :rthe*E21
_ _h, e _10 ' . s e::71:af affertyIFprem,, v"W poi.s:. __is. accep:2bie eve: , :he...eede(_bv'EN ugh :lici: j.ag : _
- :-.e v_11 se requu e:..e M
'a=*- d - %y I think that sc=e docu=enza:ic :o substa=:ia:e their clai=, bu: I do no: rec ==end analyses a: :his :1=e.
- 4. An addi:icnal though:=1ght be considered fer 11=1:1:3 the pressurize: .
dradad g. Reces: 1:ves: iga: ions f or the 205 plas:s have shown us da:
the ra:e of addi:10: ef f eedve:e has a substantial eff e:: : RC :e:perature d: p . i S.~e 'oTe.db. plan: power level only ecuir. es' aTou:"500 gy:ii.(..a: abou:
2 0-*, 0 s e c on. d s ._af : e: ::1p ) ._t..o _r e.=..ov.e...d. eca..v ..nea;;; Yet :heape= s.are . c.a..-.
. p a. -=.e p; (a: design > of abou -800'.g;= aach; vi:h reduced s:ea= genera:c: pressure the addi:1:n ra:e 1 :: eases by abou: 25* :: 30'; . ~he :c:a1 flew ra:e pessible : ends te in: oduce subcecled va:e: in:: the genera:::, fill :=
a prese: level (pessibly as a subcocied i:ven:: 7 ~ don': k::w de effe :
cf hea: pickup as de va:e: f alls :hrough the tube nes:), and : hen h.e.a:. .u p -
- b c i.l..i=. g . A :::e ;;eferab1e s. :e
.h.._e__r_e_ -e c. a. l.. .c. .-:,:...e..h...=c..e.zv_s_es.
. 4 An inves:1ea:i - in::
- .;; - -- c_aod
+' e, would rate 11=. .: in- 6 b e :: . : rod.. . i1: c. a a (valve opening res::1c:10:s, cavi.ating ven: is) say be w ::hwhile. Rate 11=1:1:3 =ay be a full c: par:ial ::adeof f f or level d:ing.
~
- 5. Further discussions vi:h TICo abou these sugges: ices are des':able; we ,
vill supper: eff e:.s in this area.
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I BABCOCK & WILCOX COMPANY
/ER GENER1 TION GROUP l
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.t. File No.
/ or Ref.
ALL
- j. .
Date
_ S'"ALL 3RIAX A.'ALYSIS *r.TE RC ?tMPS ?ctIEID -
DICIM3IR 11.1978 l
n.. i . .. . . . .. .. .
ly. csently'aoW e'vedJ h. .- -
_n .
= 1r . =~a veJ %w* 11f. -- be~
.ed.:p_.e -#mCassumind o s~f,=-aj. ,c.ide-vd
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_ove.f -
- .i.s C_S_t:12.sj A ca.c.to..r .:.:.1.- j This results 1:
actuatics of auxiliary feceva:er, and con:rol of the liquid level en :he .tecendary side of the SG a: high levels.
Should effst:e ;ever re=ais available during :be ::ansie=:, :he RC pu=ps vill re min powered and the ICS vill cen:rol the secondary side vace: level :o are cxi=a:el. R ":: E.by th
(_o.r__.a..u= ilia:.I)._ f e e d.v._a : e
_eca::ci__va_lv.e ...4. :Yu.=re-e f .e.. .S_ro:: ling :he ..=..a.in.
. - - - t=pe.r:_aaeat_.nf :he sc vate level :o =1.:.?ga. lieu, Lof small .b:eak.s,1 0S2*eands. an. analysis .cf. :he.ac. pu=ns...pouered cas4.4 1: =aking his reca. .anda:i==, ICC3 has perfe =ed a reviev :o de:er:1se what previcus ana'yses are available vi h the RC pt .ps povered a:d aise has nade an e:ginee:1_ g assess =e : ef wha: vould happes unde: this s1: a:Len.
Me pre-a.cg.jangr~14cWe Eevever, 1: is expec:ed that if a=y vero perio:=sa, they were no: dene at break si:es of i=:ere.st. Wi:h the RC pt. -
ps pevered, 1: is expec:ed tha: natural circula icn vould be main-
- ained for a lenger period of :1=a and would aide systa= depressuri a:ict.
Also, =ai=:aisance of the system flev vould decrease T he af:e reac:c: ::1p and veuld resul: in a lever pressure for systa= flashi=g :o occur. Eevever,
- .hese posi:ive influences vould be offse: by the de:: eased ability of the SG to condense stea=, folleving the less of catural cir:ulatica, due to the 1 cue: 50 level cen::al. :: is also expected : hat, vi:h :he RC pt ps ru :ing, -
a "s:ca: gecke:" vill net form in the cold legs and 1cuer e,uali:7 fluid vill ext: .h:: ugh the 1reak and thus sherten jka -d a 'a- 9J syste= :s reach the "boiline eo:" mede e# ~%e --m-e --
b* r-le-'wn facter3 at this time is I'Ehether or.ucq h A M C *D n d vi%L "a"-***Ji- the to:18d
$Ye=d ); ae earried det. i_ in_to -% e 39d undensed - r- e _ --
'--'a-'
- goya.
- atgfd%iHW -GQMJXWF L -->&-~-- A- _- - d*G Q:3srig.iif--- i.ieFaFT- +me * * .d ICOS propcses tha: :he i= pac:
of this phene =ena be exa=ined in a sensi:1vi:7 2:ud7 . ,
.,7 . - - w r.. as c As illus tra:ed_.the e . r.H*h,e *.* M, 3u, . , *ha :O . ^.= .. =*,.*E~---_3 3 -
T. running _res.u_li C2nW. . _-M Tf""* e -N*Iqd -
2.us, ICO3 rece== ends a .
. asa'.ys:.s be performed :s exa=ine this case.- _ - - IC-5 preposes tha: :his analysis be run c :he 203 7A plants fe :he fo*. levi:g reasons:
7% &
l RC Jones :o 1.1 CA :in ? age 'No Sub4ee:: S=.211 Break A. .alvs es **ith ?.C ?=e s ?evered Dec. 11. 1C73 4 i
- 1. D e presen: 205 TA s=a.11 break :opical (3A*7-10074) assu=es hea: re= oval
- ypical of :ha: for :he 177 TA pla=:s. n erefore, a co=parison of the pu=ps on case :o IA**-10074 vill pro.ide as assess =e=: of :he i= pac: c the 177 7A plants.
- 2. n ere presently exis:s an FAC :odel on the 205 TA plan s.
Bis =odel more properly accounts for SG perfo:=asce duri=g the tran- '
sient for :he 205 TA and VI?co plants. C..=parison of :he pu=ps on case to recen: 205 TA plan: studies vill previde infor=acion on the i= pac:
for 205 TA and VIPCo Plants. .
I: is expected that :his work could be scheduled in:o che ICC5 Uni: verkload now because of slippage 1: the NRC s=all break s:asdard p chle: a=d the 205 FA s=411 break work. I: is es i=ated :ha: . vill recuire 300 ch '
- ti=eIf of 3 =en:hs. :his_wo.rk_ E
- a.nd.._10.
cac: CD..C. hours in E005.for.:his issua _an. a=y questie:.s arise, please con ac:INirg._S d...a...s2am.l hi=, D2. v. u.._.L_te"Ehe' ce or call =e on e=:ension 2066. )
RCJ/lc cc: 3.1 Du=n
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- TH[ DADCQfK g # 11, CO X UMPANY ADMINI sTR A fI Vc MA.*:UAL DRAF POLICIES A!!D PROCEDURES NUMBER ~
lHPG-1707-01 (Rev 7)
SECTioN SUBJECT ,
l I
QUALITY STAllDARDS . PROCESSING OF SAFETY CONCERNS l
' I '. APPLICABILITY COMPLETE REVISION ;
ALL NPGD PERSONNEL
,II. PURPOSE To provide an orderly and visible process for identifying, evaluating and I initiating the resolution of safety concerns related to or affecting NPGD- .
supplied components, systems and services. -
To assure compliance with NRC regulations (e.g. ,10CFR21,10CFR50, etc. ).
l II. EFFECTIVITY l t
- All safety concerns identified after the issue date of this procedure.
l Safety concerns identified prior to the issue date of this procedure may be processed in accordance with this procedure.
1 IV. REFERENCES l 1
NPG-1703 Preparation and Processing of Internal Deficiency Report /
Restraint Order / Corrective Action Request i 1716-Al - Policy for Reporting of Defects and Noncompliance as Required by 10CFR21 .
V. FORMS PROCESSED (See Forms Section Manual)
BWNP-20208 - Preliminary Report of Safety Concerns (PSC)
M. GENERAL A. If guidance in interpretation of the following definition or other aspects of the procedure is needed, consult with Licensing.
B. The general definition of a safety concern as used in this procedure is:
I Any item which has been discovered during design, analysis, fabrication, l installation, testing, inspection, training, and operations activities of a nuclear power plant and which has or may have safety implications. I C. Reporting of safety concerns to the customer or the NRC is not required t if NPGD Licensing has documented evidence that the concern is adequately '
known to the affected NPGD customers in the case of potential significant I deficiencies, or is adecuately known to the NRC in the case of potential substantial safety hazards.
D. Recurrence of a previously reported safety concern shall be reported as a new safety concern. g., '
7 I REV STATL'S Kev 7 il7 7 17 l-4
@F PANES PAGF l t l .i 4h D 'l l o nie, endo a v . . ,- ==
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i THE DABCOCK & clLCOX COMPAflY 3690s.
gT ADMINisTRAllVE MANUAt.
POLICIES AllD PROCEDURES tmBER 1
l l
IfPG-1707-01 ,
1
- .. - - . . ~ , ,
I. GENERAL (cont'd) f E. Once a PSC is issued, the originator may rescind it by documenting the l basis in a memo to the Manager, Licensing, and attaching supporting docu- l mentation as necessary. The originator's manager shall indicate concur- ( I rence on the memo. Appropriate action shall be taken by the Manager, i Licensing.
(, RESP 0tlSIBILITIES FOR REPORTitlG All flPGD personnel are responsible for originating form BWflP-20208 when they ,
discover potential safety concerns that are suspected of falling within the definition given in Section VI.B. above.
- h. PROCEDURE Refer to flowchart, Exhibit A, for the procedure to process safety concerns.
i
\
-END- .
l l
. 1
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- Tile DADCOCK 6 All.COX COMP ANV
% 900 1 ADMINISTRATIVC MANUAL -
'1 DRAFT eaticies Ano eaoceouaes .EmaBER L
NPG-1707-01 , ;
v v a p EIHIBIT "A" ,
=
[
Pt0Ct5$1r.S ce gartte CC*t*te% ,
- 1 (51L 40!( 2) h 75C D I lipon discovering or receiving a reportar a potential safety I
concern suspected of falltag attnia ter arfinition of Section yt.8 l; of tne procedure prepare form SAP.22fE. "Preltainary Report m ggggggggg of Safety Concerns (P5C)*.
gg Reutew with and cttain manager's signatare on form indicating h
4 accuracy and completeness of informLter= hg s
OUAllty .
A11URAMI pgg d
5 d
l P5C <r 2 57f 8 7A j Seview for adecuacy; if necessary, rammest additional infome. s sten fron ortgsnetor. See hote 2. Atsgn PSC nu-ser ami log l gggggg; in. Prepare and distribute cover merm and inforretton espy of P5C to dtstributton listed in slote A F-- 3 Advise cultorer of poten.
stal safety concern if $3, 3 PAGJICT travtred by contract or As reoutred, convene meettag(s). reenmstog attendance from MhA0te custome* agreerent. Send the following persons:
(f bieCf!04) copy of letter ts Fanager, .)
Licenstage and Records al Cettinator of P5C i I'"I b) Licenstag Engineer asstoned to handa safety concern l c) Project f'.anagement function !
d) Ctners. as necessary, sucn as Lieursieg $ection Par.ager.
SitP 7Ar[ ,g unit manager or assignee; represetmasse of Plant Integra.
tton; Quantty Assurance.
(l _
Purpose of the meeting: i
- 4) Prerete baderstanding of concern b) Determine scaliceDtitty of concefm as operating plant and ('
plants in construction l c) Deteretne safety significance Ift? 2C TI 8) IstaDllsn priority for evalwettenstf Concern f l g ggg,ggy gggy e) SCSpe evoleatton and follow.wp acron recairt*ents (see.tted)
? including assiteent of responsseisstes and senecele. Icen- l
- f C0Vit W tify lead orgent:4 tion respopstehr for follow-up actions, f) Determine funsing reteiresients sen sources The Licenstag tagineer shall preparea sarv%ery of aesttag resvits and alstettete to atteeseesene those listed in hate 3.
g Site 7C 4 prepare follow.vo Action Cbta,in gy y,g, g3info,mation,,g g,4 g g ,, fro,m the organtasamns as necessary to
}
Plan (fAP) (see hete 1) es
$teg f, Cetain blemne Wtner of Mt W WW Mty cWM U ,
LEAD i
feDort.able er.cer f eoeral regulettent gg "5"o'e#
4;pr edt ofOF FAP from edadgers If PSC is not to be evaluated by afEC. forward saf ety concern responstbie for actions. and from PJneger. Licensing end tg cusgoner, as geer *d necessary, ses hsject rJesier functlen (Prccedgre ends Project renetta.irnt f ar.ction, alta mee to distriowlton list #6 sn ADce 3.
Destettute F AP te ledte j heee.)
Pretare evoluetton report end Ceveresse. Distrf tute in See 401444 Apr Content of report, Provtde infeemetton as accordance uttA hote 3.
necetsary to Licensthq for statedtlen report. Docupent proe,ress 45 necessary altn
- 4 flhel disposillon to Ltcensing ($tre 78) and
- distribution to those ,
listed in hate 3. N 3, l testew FAP to ettermine If !
OVAttfY furtner action is resvired A11utAhCE per h7G-l?C3-01. AJe s se .
Licenslag of action t4aen.
1P f l
gggggggg7g .f en, s.g,esie. c.,recii.s. tfg the n. maneger.
.taer cu,,f tmsLicensing.
. Review e*olvettenl report sad i
Clpe(af1 O
e
- e 9 I o., 1 (Rev 7)
7 l 7../.
. l
,.: THC BADCOCK & vill.COX CCNf'ANY I ADMINISTRA9 VE MANUAL l gpd POLICIES AND PROCEDURES NUMBER NPG-1707-01 ,
. EXHIBIT "A" (cont'd)
. a- .
- l l IVAttAll0;4 y
f( POP _f _
OA %*
1 j
MI U ~
1, 7 $ttp 7a j tacorporate corrents or resolve with art f taator. Aevise pinnaus evaluation report. If reevered. and obtata concurrence 4 g L1Cth55t G (signature on cover eve) from Managers. Ovaltty Assurance and Engineering (see note 51.
etC0eDS Ctatta Process and flie. getermine reoorting strategy. I.e.. te osom. forma t. timing. ,
etc. If tne safety concern is teentifies as at ,
- 1. Substaatta t 1s tety wa rsed fo* (ts+tswat. la r o *-u t
- e* o*
5eevices beli,eeed to aa 6G Cus'Pe* t &GCi41 resortaale).
presetly provice e.aivation report is the Civtsten aeed and appitcable Project Manager with copy to distrisutton If sted in hote 3. If written conf tenation is not ovatiagle l (VAtuA110:4 witAin 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from the custoa.ee stat the NAC nas teen 1
] StPORT nettfled, netsfy tne %AC of the safety concern by teleptore. '
TWA er telegree.
MA M - '
i v6stantial $a rety ma tard for f tst-wat. Info **c tica ce
- ices cei ne.co 'o w.3 o< 4 y w.: s er 5. r iie- '
,10UUl reportaosea, procotly provvar evaiva tion report SU61?a'atisi. gartty wa7se?g M(, to the Otetsien need and soplicable ersject Manager .ita l copy to distettetton Insted in hote J. Notify tme f.4C i 4 9 of the safety concerti witnia 48 novrs of telephone, t' .1 )
Review report. ladicate stee and cate of receist ans si n er telegree witn concurrent nottf tution to the supplier j cover rees. 8eturn copy o 'id Ih' 8F'CI'vese responsibility, signed meae to Manager. Prepare cover armo and forwrd with record of pertinent j Licensing. telepnene tpformation or copy of Tha r telegras to the
$t Itse Of FeCetal by Divt. Olvision head and dtstribution listas in note 3.
Sten Head tattiates 48- Soestt written report attitta $ says entner directly to hour reporting period tw hAC. or vta apolica61e Project tisunger to customer FW98 h d by E EII. for sutnittal to tne hAC. mita copy is destetbutton
, listed in hote 3.
- 2. jjgaf ficaat De88eicacj (10CFt50.55(el reportaale). prepare cover memo and formare wt ta evaluatm report 18 tne I
distribution listed in foote 3 and Bleesten #sead.
f vALUAT104 3. Othee pecertaele f tea. prepare cover ereo and forward 817047 witn evaivation resort to tne sistemtion listed in {
hote 3. Advise customer er hAC as vuestred.
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4 hea-tepeets61, conceca, preoare mese stattag 'eistost tion, f
- attaca e.aivauen report, and forward to distriantion Hsled la hote 3.
At itSutst3 g g 57(e 78 fesletaan PSC file active entil follow-up actions are completed.
pgg9 hottfy supeller with copy of II notification to distrl6stlen listed in lies 11.1.
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I BfPORT CettR M =
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- t. fee Setstaattel Safety Maeores. premmely motify the NU customer esta coor of aettiscation to one Oselston NM " need and enate *s. Coality Assureace preJect Manaceaent.
FUNCT104 Liceastag. Insincerlag. Costamer Seeware and sne actores Cen ter. bnere cestener nas notified see h4C. Proviae tne
- n. nager. Liceassag. setn eretten comessenties that the ABC het bete nattfled Df tne costaaer.
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p 36968'.I THE DACC0CK 6 WILCOX COMP ANY a AD'AlfJ1 STRATI vE MANUAL DMrt~y- eoticies Ano eacceouses womeen NPG-1707-01 EXH)B_IT "A" (' cont'd.)
I NOTES:
1.
All correspondence related to safety concerns shall reference file point 205/T4.4 plus the PSC number.
- 2. A safety concern that is fou'nd to duplicate the subject material of a previ-ously submitted concern shall be returned to the originator with explanation ,
and copy of the previously submitted safety concern. I i
3., DISTRIBUTION:
Originator Records Center Manager, Quality Assurance i
Manager, Licensing .
Manager, Field Engineering and Services Manager, Generic Projects Manager, Integration Managef, Engineering Manager, Plant Design Manager, Safety Analysis Affected Project Manager (s)
Other Affected Personnel, as applicable
- 4. Evaluation Report shall contain, as a minimum, the following: .
- a. Description of concern '
- b. How concern was discovered
- c. Analysis of safety considerations
- d. Equipment and plants affected '
- e. Reportability under 10CFR50.55(e) and/or 10CFR21
- f. Corrective actions, as applicable, taken or to be taken
- 5. QA Manager's concurrence indicates that the applicable NPGD organi:stions have participated in the evaluation and that an assessment has been cade to determine if changes are needed to the QA Program requirements (e.g..
. increased number of QC Surveillance inspections, increased number of vendor audits,etc.).
Engineering Manager's concurrence indicates that the evaluation report has :
been reviewed for accuracy with respect to: :
- a. Components, systems, services and plants affected
- b. Nature of the defect or failure to comply and evaluation of the safety concern
- c. Corrective action .
t e
e
- I
+
THt saecock a witcoX COMP ANY fg..i.
{
POL i ES AND ROC DURES 7 T NUMBER NPG-1707-01 !
EXHIBIT "A" y
(cont'd.)
y
~T NOTES: (cont'd.)
- 6. Follow-up action is defined, for purposes of this procedure, as follows:
- a. To bring deficient items,into conformity with requirements
- b. To identify causes for deficiency
- c. To prevent recurrences of deficiency
- d. To make such other investigations or analyses or take such other follow-up actions as are deemed necessary because of the repeated concern.
- 7. The follow-up action plan shall include as applicable:
- a. Actions to be taken
- b. Individuals or organizations responsible
- c. Schedule for completion including milestones
- d. Deci; ion points and alternate actions
- e. Funding source t
e k a e
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"'** C 1 0 .,, gy y 11-20-79
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M A {r--l BABCCCK & WILCCX 87sP-20204-2 (11 73)
PRELIMINARY REPORT OF SAFETY CONCERNS
,lTO: m Acta. UctnstlG, NPG3 C0:3 ULT sPC0 Lactasise L*' 88-(CC QUALITY Assutaagt) l
, FOR A3313famCE as 74 4 ,
usem Fsa se. :
COMPLift:0 TNi$ 704N C0sitACT ao.
Dasas12Aftos:
ATTACR A40 10fsflFY. 37 #ACf MUMatt. ANY SUPP04Tisc saf0tMAT10s/00CUMENTS gjrats.s0rAa0OsuniCNPLANTWA3THE3AFETYConcias 3.lfD Tout t#0st10CE IS CUSTONES AWARET CYE3C X0 fetsfLF8107 . I wars a sow ,
l 4j 10 TOUR K A0rLIOGE .IS N AC Atatt? QYt3 O A0 vars a mov 1
l lj Ofs(A AFF(Cf(0 CONTRACT 3 (CUST0N(A AANE AND LOCAT40s)
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I sj0(3CtlPflosCF 3AFITY C0aCERA.40tNTIFY AFFECT 10 ComPostaf(3), 3Y3TIW(3) 02 ACTaflTTf3UPPLIER. As0 IMPACT Os I SAFiff CF PLA8T OP!!ATICES l
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