NRC-85-3061, Submits Internal Review Results Concluding That Safety Deficiencies,Intimidation & Retaliation Alleged by Ja Segletes Did Not Occur,Confirming 850712 Preliminary Assessment.Related Info Encl
| ML20198M834 | |
| Person / Time | |
|---|---|
| Site: | Indian Point, Turkey Point, Diablo Canyon, Comanche Peak, 05000000 |
| Issue date: | 09/13/1985 |
| From: | Rahe E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Taylor J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| Shared Package | |
| ML20198M785 | List:
|
| References | |
| FOIA-85-654 NS-NRC-85-3061, NUDOCS 8606050434 | |
| Download: ML20198M834 (25) | |
Text
.
- 3
'N s d I l
Westinghouse Water Reactor Ba 32 Electric Corporation Divisions "'mePuma n2 man September 13, 1985 t
NS-NRC-85-3061 Mr. James M. Taylor, Director Office of Inspection and Enforcement -
U. S. Nuclear Regulatory Comission Washington, D.C. 20555
Dear Mr. Taylor:
O .
On July 12, 1985, I wrote to you confirming our telephone conversation with you concerning allegations raised by Mr. John A. Segletes, a former employee of Westinghouse Water Reactor Divisions (WRD) at the Monroeville Nuclear Center.
At that time, we informed you of our ongoing internal review and that our preliminary review had revealed no actua], safety deficiencies as a result of the alleged incidents and practices. We also had found no reason to believe that Mr. Segletes or any other employee had been inhibited from raising safety concerns through the established channels as defined in the com;mny's plicies and procedures.
We have now completed our internal review of all the matters raised by Mr. Segletes and the results of our internal review confirm that our preliminary assessments were correct.
~
I
'Ihe matters potentially related to safety were reviewed by persons technically competent in the areas involved and by members of the Nuclear Technology j Division (NTD) Quality Assurance Department. The results of their review '
including recomended corrective actions were documented and presented for consideration by the Water Reactor Divisions Safety Review Comittee in accordance with WRD policies and procedures. The results of this review process are stenarized in Attachnent.l.
Mr. Segletes' claims that he was inhibited from raising safety concerns were investigated by a Select Comittee reporting to the General Manager of NTD.
The Select Comittee concluded in their report to the Division General Manager -
as follows:
"The comittee finds that the allegations of Mr. Segletes on intimidation and retaliation are not supported by the evidence and that a more plausible reason for the allegations may be found in the employee history and ultimate termination of employment, not from safety concerns." l M
B606050434 860527 PDR FDIA DOHERTY85-654 PDR (
NS-NRC-E-3061 September 13, 195 Page 2 The results of our internal retiews were shared with members of your staff during the inspection concerning this matter conducted during the week i
beginning on Monday, August 26,195 and have been conmmicated, as '
appropriate, to affected licensees. We will respond to the staff's report formally when it is received.
Fiaase call me (412 - 374 4868) if I can be of fbrther assistance. ,
Very truly yours, !
E. P. Rahe, Jr. Anager Nuclear Safety artment MPO:pj .
Attachment e
e O
m E
ATTACHENT 1 i Af1 mATTm NO. 1r i nRT nTART_n CANYON MAETY ANat YRrR Ris allegation states that "Almost all of the calculation notes that support the Diablo Canyon Final Safety Aralysis Report are missing" and that they "have never been retrieved. ..."
Remnits of Reviewt .
Bere are no missing analyses nor are there required analyses which were never j
perfomed. The microfiche copy of one analysis was found to be missing and was re-filmed from the hard copy which was not lost. An independent quality ' i t )
! assurance review of the records confirm that the information needed to support '
the current Diablo Canyon analysis basis is readily available for review and is appropriately protected by microfilming. Thus there is no safety significance -
attributable to this allegation and no further action is required.
Aff MATTON NO. 2r NONDIStifEURE OF AN UNRAq cnNnITION AND RETALTATTON BY MANE ER This allegation states that a modification of the turbine runbeck system (deleting the flux rate signal) in Turkey Points Units 3 and 4 would violate .
IEEE-279, the single failure criterion, and redundancy requirements.
Ramn1 ts of Reviewt The modified turbine rtnbr.ck system as finally implemented in Turkey Point Units 3 and 4 satisfies the requirements of the licensing basis for those plants. Westinghouse initially was requested to perform an analysis to determine if the system could meet the controlling licensing criterion (DNB ratio > limit value) if the flux rate signal used in the original design were deleted.
The analysis showed that the criterion could be met. Based on this analysis, FP8L obtained NRC approval to delete the flux rate signal.
Subsequently, FP&L asked Westinghouse to review the actual modification of the system. i As a result of that review, Westinghouse found that the modification would not provide the degree of diversity that was provided in the original design and recommended a change that would restore the original diversity.
None NRC.
of this in any way affected the validity of the analysis approved by the FP8L has implemented the design change recommended by Westinghouse and the system satisfies the NRC licensing requirements for Turkey Point Units 3 and 4. Thus the safety of the Turkey Point Units is assured and no further action is required.
All m& TION lEL ~4: FFAR OF RETAI TATTnN This allegation raised the same issues as were raised by allegation No. 2 in connection with a similar analysis performed by Mr. Segletes for Indian Point Unit No. 2.
Ramn1ts of Review Consolidated Edison Company has not yet submitted the subject analysis to the NRC for review and approval nor has it made any modifications to the turbine rtmback system. Thus there is no adverse effect on the strety of Indian Point i Unit No. 2.
l
- l l
Results of Review (continued)
When the review of the modification of the Turkey Point turbine rmbeck system
' modification was done, Mr. Segletes was instructed to prepare a letter to be sent to Consolidated Edison Company similar to the one sent to FP&L to advise '
l l them of the need to maintain the diversity provided by the original design and to recoasnend a way to do it. The letter had not yet been prepared when Mr.
Segletes left Westinghouse. Subsequent to his leaving, the letter was prepared and sent. During the NRC inspection related to this matter, a discrepancy was fomd between the descriptions of the turbine rmback system in Chapters 7 and
, 14 of the FSAR. It was verified with personnel at the plant site that the l description in Chapter 7 is correct. Since the letter was based on the description in Chapter 14, the details do not reflect the plant configuration. ,
However, the conclusions remain valid. A correction has been sent to the Consolidated Edison Con:pany. No further action is required.
Af f BY:ATTON lEL 41 ANAf YMFA THAT RATMF THE RFACTOR WTP (El TURRTkF HIP SEIPOINT . . e .
This allegation states that an error was discovered in an analysis performed to support the deletion of reactor trip on turbine trip fmetion at or below the P-9 permissive setpoint (corresponding to 50 percent of full power) for Comanche Peak and that several other analyses may have been performed in a non-conservative sanner. The error concerned the modelling of reactor coolant inlet temperature in the calculation of the DNB ratio.
Results of Review The analyses referred to in the allegation are.P-9 analyses which are similar
~ to the analysis performed for Comanche Peak. These analyses employed an '
assmption that reactor coolant flow is maintained until just prior to the time of reactor trip by a turbine generator motoring feature which maintains power to the reactor coolant pumps for approximately 30 seconds after turbine trip.
Similar analyses had been performed assuming a more conservative assumption that there was no motoring and hence earlier loss of reactor coolant now and lower DNB ratios. Actually, the latter asstaption is mnecessarily conservative since the loss of now before reactor trip could result only from multiple failures. The reduction in DNB ratio resulting from the more conservative assumption of no motoring is small in relation to the high minimtm DNB ratio which occurs as a result of the low power level from which this transient is initiated. The limiting transient DNB ratio for loss of now occurs in the loss of now transient from full power. ,Because the results of j
this transient satisfy the controlling licensing criterion (DNB ration-> limit -
value), there is no safety concern involved.
The P-9 analysis is a combination of loss of load and loss of flow transients and standard procedures for these analyses are used in performing P-9 analyses. Specific guidance for P-9 analyses is being prepared for incorporation in the standard procedure for performing the loss of load i analysis. i
, , _ _ _ _ , , -,,,r -w,, - =-rv- '~'*w"'^~ ' " ' ~ ' " " ' " ~ ~~ ~
i i
I AI_f EY'.ATTnN 4Br DROPPED ROD ANAT YRFR FDR 'mRETNF RUNRArlf PI ANTS This allegation states that an error in the analysis methodology for plants using the turtine rmbeck feature for mitigation of the dropped rod transient was incorporated in several analyses and that the analyses were not corrected when the error was found. ,
Renn1ts of Review At the time the error (incorrect mcertainty on the turbine runbeck setpoint) was discovered, it was detemined that sufficient DNB margin existed to ensure safety of affected plants pending re-analyses of the transient. It was also I
determined that the results of the error were nor> conservative only for those cases in which a reactor trip occurs via an overtemperature delta-T signals.
i Re-analyses have been completed which show that the DNB design basis is met for the current fuel cycles. Verification that the DNB design basis is met for future process.
cycles will be done as part of the standard reload safety evaluation The standard procedure for performing these analyses ha.s been revised .
to incorporate the correct mcertainty of the turbine rmbeck setpoint. l I
Ali MATION NO. 8s; NM-UNTEDRM PROrFillfRFR WITHIN WESTTNnHnffRF FDR RFpoin iNC '
PQTEnlTTAT. RAhi1 VTor ATTnMR This allegation stated that Westinghouse internal procedures were inconsistent concerning reporting of potential safety issues.
Rann1 ts of Evaluation This on plant allegation safety. didThe not raise any safety concern and hence there is no effect merits of the allegation were addressed by the select '
~ coamittee which found no basis ft:r the allegation. Nevertheless, to ensure that no one has any mismderstanding, the Instruction and Guidance for reporting potential safety issues is being revised to emphasize that any of the channels for reporting to the Safety Review Coneittee or to any level of management are always open to anyone wishing to raise a safety issue.
AlJ MATION No. 6At TRANSMITTAT. OF PRFf TNTNARY DRAET RFSPNSFR This allegation states that the factional requirements for the Italian 1
! reference plant were not marked preliminary, were not reviewed in-house, and )
were procedures. not approved by a manager and thus did not conform to Quality Assurance l l
Results of Review I
'Ih'e the NRC.
matter raised by this allegation did not involve any U.S. plant licensed by The purchase order for the work was issued by our Italian Licensee and called for preparation of preliminary fWetional requirements and did not call for our in-house review and management approval. Also the transmittal letter identified the preliminary nature of the docments. . Thus it is clear that there was little, if any, possibility for the licensee to mistake the documents for final ones. Therefore, there are' no safety implications resulting from this allegation.
i
af f rn1TinN 6Br AMIONTNG COMPETENT INDEPENDENT VERIFTERR his allegation states that Risk Assessment Technology section procedures for assigning calculation note reviewers are " rarely if ever followed",. i.e., that managers do not assign reviewers.
1 Results of' Review Docmentation that an engineer is qualified to perform reviews is contained in the Quality Assurance Training records in the group. These are made available together with an analysis checklist for the engineers to use in getting qualified reviewers to verify their work. As a result of recommendations from previous audits, a coenitment had already been made to update and clarify procedures.
In addition, the analysis checklist has been amended to note periodic management review and approval. This does not change the process but, rather, it ensures that it is docmented. A review of all calculation notes written sinae September 1,1984 has not revealed any instance of the use of, unqualified reviewers as verifiers of the work. Thus there is no safety' ~ ~ -
concern connected with this allegation.
All mATION NO. 7! THREATENED RETAI TATION FDR SENDTMC WRi t itM WMAnES This allegation alleges harrassment by Mr. Segletes' manager.
Remits of Review No safety concerns were raised in connection with this allegation. The select comittee found no basis for this allegation. Thus there are no safety concerns arising from this allegation. .
~
ALL EATION NO. 8; POOR CAI CULATION NOTE CHErI(TNG WHICH RESULTED IN OUAI TTY AEllRANG VTOf ATIONS AND NON-CONSFRVATIVE COMP [nEM INRJT DATA This allegation lists " errors" found in a review of a calculation rete and states that the reviewer was not qualified.
Results of Review According to the guidelines in use for the selection of qualified reviewers (see the response to Allegation 6B), the reviewer of this calculation note was qualified.
However, the calculation note was re-reviewed by a different qualified reviewer.
The " errors" noted by Mr. Segletes can be characterized as either technical errors or adDinistrative errors. Administrative errors (e.g., page not numbered) do not affect the input, results, or conclusions of the i calculations. Consequently, there is no impact on plant safety due to these errors.
The of thetechnical errors noted also have no impact on the results and conclusions analysis. This is because they refer to' input parameters that are not actually used in this particular analysis. The computer code used for this analysis has many options. If a particular option is not exercised, then the
___ . . - . . , - - - - ~ ~ - - - ~ ' ' ~ - - ~ ' ' ~ ~ ~ ~ ' ~' ' ^ ^ ' ~ ^
Remit ts of Review (continued) input supporting the use of option becomes irrelevant ar.'d need not be reviewed. The correct input for the code and options are contained in the procedures for performing the analysis and in the computer code information manuals. These are used by both the analyst and reviewer. Therefore, while it ,.
may appear that there are input- errors, in fact the input in question was not
- used by the computer code and thus had no impact on the results. Therefore, there is no impact on plant safety.
4 e
O Y
l
_ , . . _ _ _ . . . _ . _ . . . _ _ . . _ _ _ _ _ _ _ _ __ _ . . _ _ _ _ _ _ . _ , _ _ . _ _ _ . _ , . _ _. __ _ ._ : : a. . _ _ . . . _ _.
~u 2
- s FROM:
NS-85-1366 NUCLEAR SAFETY DEPARTMENT
- h EXT. : 284-4858 DATE: August 9,1985 alBJECT: SAFETY REVIEW COMMITTEE EETING T3: K. F. Cooper B. L.. King /T. J. Mitlo/R. A. Stokes l L. B. Kincaid/E. F. Duhn S. H. Kale /D. R. Grain 1
- 0. J. Woodruff W. D. Fletcher '
F. L. Langford H. F. Menke l A. E. Blanchard R. A. Wiesemann ec: C. H.Alsing Hammond [ h Y N P. T. McManus D. N. J. L. Little l P. A. Loftus M. P. Osborrie D. C. Richardson E. K. Figenbeun [ n%
J. L. Gallagher R. Saint Paul ed!7 NC-A. F. Phillips J. D. Campbell < 7 4(/g [lb.,
A. A. Fesler T.W.T. Burnett c.k N. * ?,e Q[%g W. Stevens E. M. Burns fj R. E. Lowder P. R. Delhaye ..( ' ,.
j J. Cohen .y . , .,
u I
The Safety Review Cotttee will meet to review certain of the allegations contained in the attached letter from Mr. John Segletes, a fonner Westinghouse Nuclear Safety employee. The allegations regarding intimidation and retaliation are being addressed by a separate Select Connittee, reporting to Mr. Gallagher and will not be addressed in our meeting. We will heview the
, renaining allegations, which can be categorized as technical and procetral. -
'Ihe technical presenter will be Melita Osborne. The proce@ral presenter will be Dave Alsing.
In view of the traique nature of this situation, I would like to have a slightly different process than we nonnally follow. F1 rst, I would like each of you, after reading the material to sutznit any " top of the head" questions in advance (by August 14, or as soon thereafter as possible) to Mr. E. Figenbeun. This will allow the presenters to be more fully prepared at the meeting and will, hopefully, make the meef,ing more efficient. Ariy and all additional questions, of course, will be handled in the meeting itself.
Secondly, I have asked the presenters to describe any a'ctions that are being .
taken or planned on the basis of the results of the review of Mr. Segletes' allegations. After our decision on reportability for each allegation, I will ask the Committee for their evaluation of the actions being taken or planned together with any consensus reconnendations it chooses to make in addition to !
its reportability decieion. This portion of the meeting will therefore be somewhat equivalent to a Design Review-a rule that the Connittee has traditionally avoided.
L \\
\
I-85-1366 August 9, 1985 )
Page 2 1
l l
Because of the length and complexity of* the allegations, we will start the session at 10:30 A.M., and have a working Itach. The meeting will be held on August 27,1985 in Conference Ec m 415 at the Nuclear Center. '
Again, I urge you to read the material in advance, subnit questions, and above all, attend the meeting.
/ O
. P. Rahe, Jr.
Chainnan WRD Safety Review Committee p
Attachment -
o O
W e
i
- 2. a.
?-
C ~ ~ - .
NS-OR.S-OPA-85-266
_ ~%
NS-RAT-PTA-$-216 FROM: Nuclear Safety Department i WIN: 284-4901/284-4481 f DATE: July 15,195 .
- ~-
SJBJECT: Request for PI on J. Segletes Letter to NRC -
Keywords: PI/SRC/QA/ QUALITY ASSURANCE PROCEDJRES/SEGLETES/NRC 70: R. A. Wiesemann MC 4-01 cc: E. P. Rahe, Jr. MC 4-12 W. J. Johnson HNC 4-16 J. D. McAdoo MC 4-12 J. L. Little MNC 4-17 D. C. Richardson Expo Mart D. N. Alsing MNC 3-10 E. K. Figenbeun MC 4-01 B. Cantineau WNI-Brussels R. P. DiPiazza MC 4-08 OPA/PTA Personnel
~
C. H. Hammond Penn Center - .
On July 11, 1985 Westinghouse received a copy of a June 17,195 letter from J.
Segletes to the Director of Inspection and Enforcement at the Nuclear Regulatory Comission. In the letter (Attactynent 1), Mr. Segletes indicated that he had either been involved in or had knowledge of incidents that he believes are violations of either nuclear regulatory law or Westinghouse quality assurance requirements. Of the eight issues raised by Mr. Segletes, four of the issues allege potential non-conservatisns or errors in the current safetj analysis of the following plants: Indian Point 2 & 3, Turkey Point 3 &
4, Comanche Peak 1 & 2, Point Beach 1 & 2, Ginna, Beznau, and V. C. Sunmer.
. One issue is concerned with missing safety analysis docunentation for Diablo Canyon 1 & 2. The remaining three issues are concerned with the content, irt.erpretation, and implanentation of the Westinghouse Quality Assurance
, Program and the WRD Operating Procedares for reporting potential safety Concerns.
On July ll,1985, Nuclear Safety informed the NRC that we had just become aware of the 6/17/$ letter and that we had initiated an internal review of the factual matters contained in Mr. Segletes's letter. The formal ccannunication follm#ing up the 7/11/$ telephone call is contained in Attachnent 2.
Our prelimirary review has revealed no actual safety deficiencies as a result of the alleged incidents and practices. Supplanental infonnation supporting the safety related issues along with actions currently imderway is provided in Attactanent 3. '
, l I J
',e, EOR.S-OPA-85-266 ERAT-PTA-85-216 July 15,1985 Page 2 Operating Plant Analysis and Plant Transient Analysis reconmend that a Potential Item raaber be issued to M. P. Osborne, Manager, Plant Transient Analysis and P. A. Loftus, Manager, Operating Plant Analysis. The managers of OPA and PTA have the responsibility for determining the safety significance of the issues identified. D. N. Alsing, Manager, Quality Assurance Systems and :
Compliance along with the OPA and FTA managers have the responsibility to address the quality assurance and proced.iral issues. i l
k/Yflil P. A. Loftus, Manager Operating Plant Analysis M. P. Osborne, Panager Plant Transient Analysis
/pj J Attachment 6,
e 4
- D
. _ . - . - - - . - , . - - - , . - - - - --,-.---,,-,-n,, - -
.n. - --.- .,-, - --
k,ik, Vg gr.
C.
~
Attactment 2 Westinghouse Elecinc Corporation Water Reactor . uw nw::n: cm Divisions gg ammoc..:, t a ta
~
G -t m.q.- n' - % 'M.
, July 12, 1985 !
Mr. James M. Taylor, Director Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Taylor:
This is to confirm our telephone conversation of July. 11,. 1985-in which we informed you that on that date we first becane' aware of a letter dated June 17, 1985 and sent to your office -
by Mr. John A. Segletes , a former employee of Westinghouse Water Reactor Divisions at the Monroeville Nuclear Center.
Mr. Segletes alleges in this letter certain practices and events in which he was involved while an employee which he believes were violations of either " nuclear regulatory law or Westinghouse quality assurance requirements", and requests your investigation and appropriate action.
Please be. assured that we shall cooperate fully in this regard,
' and that we have begun an internal review of the factual matters '
contained in Mr. Segletes' allegations.
Our preliminary review has revealed no actual safety deficiencies as a result of the alleged incidents and practices. Moreover, we have found no reason to believe that Mr. Segletes or any other enployee has been inhibited from raising safety concerns through the established channels as defined in the company's policies and procedures. Our continuing review will place highest priority on verification of the safety of licensed facilities, and our findings will be communicated as appropriate to affected licensees and to your office.
Ple'ase at call any time. me (412-374-4868) if I can be of#further assistance Very truly yours, I
.k w.,
r P. Rahe, Jr., Manager -
'uclear Safety Department-1 W JY *
. n Ivpa u / a' C,]e
. l l
ATTACHMEN1 3 l l
l l
- 1. LOST DIABLO CANYON SAFETY ANALYSES '
Background on the Allegation t he allcoati on states that "almost all of the calculation notes that support the Diablo Canyon Final Safety Analysis Report are missing." and that they "have never been retrieved...".
Status and Safety Impact !
Transient Analysis has reviewed the files for the Diablo Canyon FSAR calc i notes that form the current licensing basis for the FSAR non-LOCA safety analysis (including both Chapters 6 and 15). These analyses and the corresponding calc notes are listed on the followinq pages. There a're no l missing analyses nor are there accidents for which an analysis was never performed. The microfiche copy of CN-RPA-74-73 (Feedline Break), which is the official, QA record, was found to be missing and has been re-filmed from the paper copy, which was not lost. No analyses need to be recreated due to lost files. There is no impact on plant safety due to this issue.
Action to Resej,e.
None required. -
O 9
0
ATTACHMENT DIABLO CANYON SAFETY ANALYSES EVENT CURRENT ANALYSIS BASIS
- 1. Rod Withdrawal From Subcritical CN-TA-84-119 Rev 1, PEG /PGE Rod Withdrawal From Suberitical
- 2. Rod Withdrawal From Power CN-RPA-75-45, PGE/ PEG 17x17 Amendment Chapt. 15 Accident Analyses *
~
- 3. Dropped Rod PGD-83-323, PGE and PEG Cycle 1 Rod Drop Analyses
. -s ..
WCAP-10297-P-A,' Dropped Rod Methodology For Negati ve Flu:t- -
Rate Trip Plants
- 5. Partial Loss of Flow CN-RPA-74-116, PEG (17x17)
Loss of Flow / Locked' Rotor
.6. Complete Loss of Flow CN-RPA-74-116, FEG (17x17) .
Loss of Flow / Locked Rotor CN-RPA-74-62, Underfrequency Transient - PEG and PGE
- 7. Locked Rotor CN-RPA-7^-116, PEG (17x17)
Loss of Flow / Locked Rotor
- 8. Startup of Inactive Loop
- CN-SA-AA-73-191, Excess Feedwater Malfunction and Startup of Inactive Loop For 17x17 ,3 and 4 Loop Plants
- - Unless noted in Event column, current analysis is for both units.
e
.j
EyENT _
CURRENT ANALYSIS BASIS 9 Feedwater Malfunction CN-SA-AA-73-191, E:: cess Feedwater Malfunction e n ti Startup of Inactive Loop For 17x 17 3 and 4 Loop Plants CN-SA-AA-73-204, Correction to F.W. Temperature Drop Due to Bypass Valve Failure as Reported in CN-SA-AA-73-191
- 10. Loss of Load / Turbine Trip CN ,RPA-75-45, PGE/ PEG 17x17 Amendment Chapt. 15 Accident Analyses
- 11. Loss of Normal Feedwater CN-RPA-75-45, PGE/ PEG 17::17 Amendment Chapt. 15 Accident Analyses .
.s .
- 12. Station Blackout CN-RPA-75-45, PGE/ PEG 17x17 Amendment Chapt. 15 Accident Analyses
- 13. Excessive Lead Increase CN-RPA-75-45, PGE/ PEG 17::17 Amendment Chapt. 15 Accident Analyses 14 RCS'Depressuri:ation .
CN-RPA-75-45, PGE/ PEG 17x17 Amendment Chapt. 15 Accident Analyses -
- 15. Credible Steamline Break-PGE # CN-RPA-78-239, PGE Credible Steam Break (Lower Pr:r.
Pressure Setpoint)
- 16. Credible Steamline Break-PEG # CN-RPA-74-87 Rev. 1, Documentation of the PEG and PNJ steambreak analyses for the 17x17 FSAR amendment
- - Analysis under PI has been review, opened as result of setpoint[ inconsistency.
(PI-85-012) 1 e
e
EVENT CURRENT _ ANALYSIS BASIS _
17 Spurious Safety Injection CN-RPA-74-49, Spurious Operation o+ the Safety Injection System at Power
[17x17 Amendment]
- 18. Small Steamline Break (Bounded by Large SLB)
- 19. Large Steamline Break / Core Resp.-PGE CN-RPA-74-70 Documentation of the PGE Steam Break Analyses for the 17n17 FSAR Amendment CN-RPA-76-1, PGE Steambreak With Single Feedwater Valve Failure and Maximum SIS Flow _
- 20. Large Steamline Break / Core Resp.-PEG CN-RPA-74-87, Documentation o f. .
the PEG and PNJ Steambreak Analyses for the 17x17 FSAR Amendment
- 21. Large Steamline Break / Mass & Energy CN-RPA-78-202, Documentation of PGE/PNJ Containment Analyses Following Main Steamline Breaks 22.'Feedline Break # CN-RPA-74-73, Feedline Rupturo Analysis For PEG 17x17 Amendment
- 23. Rod Ejection ND-III-531, Red Ejection Parameters for PGE/FEG Cycle 1 CN-RPA-74-48, Reanalysis for POR 17x17, Including Effects of Fuel Densification CEOL-HZP3 CN-RPA-74-36, Rod Ejection Transient Analysis for 17x17, Including Effects of Fuel Densification CBOL-HZP,EOL-HFP,BOL-HFPJ O - A reanalysis of feedline break event is underway to address pressuri:er liquid relief issue.
-,-- m m a a m, - --
i l
l
~
- 2.' NONDISCLOSURE OF AN UNSAFE PLANT CONDITION AND RETALIATION BY MANAGER Background on the. Allegation the turbine runback system provides protection for a dropped rod event by reducing turbine load to a preset power level. In the standard desian, ,
the turbine runback is antiated by either a rod-on-botton or 1 out-of 4 neaative flux rate signal. Ihe allocation states that the proposed modification (deletano the +1ux rate signal) violates the requirements of IEEE-2/9 one of which is the sinole failure criterion, and that the proposed change creates a system which as not redundant.
The turbine runback plants are an older category of plants for which the licensing bases do not meet current licensing criteria, but which were reviewed and approved by the NRC at the time the licenses were issued.
i The turbine runback system performs a safety function in that it must be actuated followinq a dropped rod / dropped bank in order to show that the licensing criterion ( DNBR >1 i mi t value) is met. However, the original design of the turbine runback system does not meet all the criteria of.
IEEE-279, is not fully redundant, is not environmentally or seismically 1 qualified, and is not Class lE.
~
Current protection systems meet these requirements. The system design changes were developed to be consistent with the licensing basis (as opposed to requiring the. system to meet s current criteria). The fact that the modified system al'so does not meet all of the above requirements does not violate the licensing basis of the plant. This was the context of the discussions between Mr. Seglet.es and Ms. Osborne.
1' -
Status and . Saf ety_ Imp _ac_t FP&L submitted the analyses performed by Westinghouse to the NRC for review and approval to delete the flux rate input to the turbine runback system. The NRC approved the modification in an SER dated January 6, 1983. In February 1985, FP&L had recently implemented a hardware change on the Unit 3 turbine runback system which deleted the flun rate input and asked Westinghcuse to review the hardware modifications. (The original i
proposal had only addressed analyses, not hardware changes.) They
_ specifically asked about the single failure criterion. A summary of the Westinghouse response (NS-RAT-PTA-85-091) is that the new system must meet the criteria to the extent that the old system met the criteria. Certain portions of the system meet the single f ailure criterion, but the single failure of the turbine does not meet the criterion. Westinghouse suggested a means by which the implementation could be done. FP&L undid their modification (i . e. , re-instated the flux rate input) at that time.
They did not file a LER because*their internal review concluded that none was needed. FP&L plans to implement the revised hardware design in the upcoming refueling oQtages at Turkey Point 3&4. The current analysis j approved by the NRC remains valid and there is no impact on the safety of the plant. .
Action to Resolve a None required.
e
- 3. FEAR OF RETALIA~lIUN Backaround Mr. Segictus performed a similar analysts for Indian Fotnt 2 such as had been pertermed for FP&L's lurkey Point units. When FP&L asked questions related to the implementation of the modification, Mr. Sogletes was told by Ms. Osborne to write a similar letter to Consolidated Edison bo+cre they implemented their change in order to ensure that the issue would not arise tar Indian Point.
Statu.s and,.Saf ety _ I,mp.act, ,
The letter had not been written when Mr. Segletes left Westinghouse. The action has been reassigned to G._Heberle and J. Polavarapu. Consolidated Edison has not yet submitted the analysis to the NRC for review nor have they made any modifications to the turbine runback system (since NRC -
approval has not been received). Thus, there is no safety impact on the plant.
Ac t,i on to_ Resolve ,,
Issue a letter to Consolidated Edi non to clarify the i mpl eme'n t a t i on of the turbine runback modification. This will be done by 7-31-85. (Osborne)ff-L aTTe ra. N 5 IRAT- PTq 85-Z'58 ATT AO+ E,D 9
- =
1
l
.f To: CANTINEAU.B (75:WEU102)
Cc: LITTLE. J.L (WST6162)
Cc: OSBORNE.M. P (WST8665)
Frcm: OSBORNE.M. P (WST8665) Posted: Wed 24-July-85 14:01 EDT Sys 49 (13) ,
Subject:
BEZNAU DROPPED ROD ANALYSIS - SEGLETES LETTER I IN EE SEGLETES LETTER, BEZNAU IS LISTED AS CNE OF THE FLANTS FOR WHICH THE DROPPED RCD ANALYSIS HAS A NCN-CONSERVATIVE 11JRBINE RUNBACK ERROR. A POSITIVE UNCERTAINTY IS CONSERVATIVE. HCMEVER, WE ERROR ONLY APPLIES FOR DROPPED ROD CASES WHICH TRIP ON OVERTEMPERAWRE DELTA-T. OTHERWISE, IT HAS NO IMPACT.
WE CALLED DAN RISHER CN HONDAY AND HE 10LD US THAT THE BEZNAU ANALYSIS FOR DROPPED ROD HAS THE ERROR BUT NEVER TRIPS CN OVERTEMPERAT1JRE DELTA-T. IF SO, THEN NO REANALYSIS IS NEEDED. PLEASE CONFIRM THS BY E-P/IL AS SOCN AS POSSIBLE SO IT CAN BE INCLUDED IN THE PI FILE.
REDARDS, ELITA CSBORNE ,
o To: 42L Cc: OSBORNE.M.P (49:WST8665)
CANTINEAU.B (WEU102)
From: CANTINEAU.B (WEU102) Posted:
Subject:
Thu 25-July-85 NOK DROPPED ROD ANALYSIS 14:50 BST BRITAIN INT.REF. SA/85/564 .
INTERNAL COPY: D.RISHER
~
IN REPLY TO YOUR E-MAIL OF I CONFIRM THAT OVERTEMPERATURE DELTA-T REAC FOR ANY CASE ANALYZED, SPURIOUS RUNBACK) . THIS MEANS THAT THE DIR RUNBACK SETPOINT ERROR HAS NO IMPACT ON THE ANALYSIS.
COULD YOU PLEASE SEND US A COPY OF THE SEGLETES LETTER .
REGARDS BOB PRIOR SYSTEMS ANALYSIS, BRUSSELS ,
l ch l
l
' I
a l
- 5. NONUNIFORM PROCEDURES WITHIN WESTINGHOUSE FOR REPORTING POTENTIAL l SAFETY VIOLATIONS l l
Background l The two letters referenced in the allegation (NS-RAT-PTA-85-047 and NS-RAT-PTA-85-051/NS-OPLS-OPA-85-046) are attached. The " undated memo" referenced in the allegation is attached to the first letter. l S t a t_us , an d _Sa f et y.. .I mp_ac t PTA and OPA do not feel that the procedures are inconsistent as discussed in the attached letter replying to Mr. Segletes' request for clarification. There is no impact on plant safety.
Action to Resolve None required.
l s
O O
O
\ .
S
- zk NS-TA-83-520 SATO-CSDA-450
% NTD we 284-4866/4481 '
are December 16. 1983 mIRP EITNCTIONAL REQUIRDENTS REV O to H.J. Fix
- International Licensing PC 3/600 3-29 cc: N.P. Mueller w/o att -
^ '
J.L. Little w/o att A. De St Maurice w/o att ~
L.I. Ortenberg w/o att Revision 0 of the Italian Reference plant Control and Protection System Functional Requirements are attached. These should be transmitted to HIP?-,50PREN as part of the effort on Performance Order #10. " Functional D( '.gn Documents."
r These documents are issued with a " Preliminary" status per WRD procedures
( to allou NIRA/SOPREN the opportunity to review and comunent. NIRA/SOPREN
- may transmit these documents to the Architect Engineer and/or utility for
- review and comment if desired.
These documents are working documents which are periodically revised to reflect evolving plant design. The second external issue. Revision 1 and all succeeding issues will be " approved."
Attachment II provides some clarifying comunents on the Protection System Functional Requirements. Attachgent III does the same for the Control System Functional Requirements.
If you have any questions, please call G.D. Storrick for questions in the Control area and for general administrative questions. and M.P. Osborne for questions in the Protection area. .
, j:, -
LAA.- Q C/D. Storrick M. P. Osborne
' Control Systems Design & An& lysis Transient Analysis Isag att (
Osca00ce:
B. Assigning a Competent Independent Verifter Within the Risk Assessment Technolocy (RAF) Section Background on the Allegation the allcqation states that the RAI procedures for asstaning calc note reviewers are " rarely 1 4 ever followed" i.e., that manaaers do not assion
+ reviewers.
Status ynd Safet) ,1m uct the F.Ar p.ocedure F.ates that a manager will assi gn a revi ewer but does not specify axae*.Ay how this is to be done. If revtew responsibility is not assigned directly by the manager, PTA and CPA policy is that encaneers may select their own independent and qualified reviewer. Documentation that an engineer is qualified to perfqrm reviews is contained in the UA training records in the group. These are available for the engineers.to review in selecting a qualified reviewer for a calculation. There is not, however, a formal periodic review by management of the files to verify qu.nlification of reviewers. This was also noted in the internal Nuclear i Safety audit (letter PA-85-850, dated 5-22-85) and a recommendation was written concerning this. As a result of this recommendation from this.and another audit by Southern Company Services, PTA and OPA had already committed to update and clarify this procedure by October 31, 1985 ( l e t't er NS-RAT-PTA-95-201, dated 6-26-85).
All calc notes written since September 1, 1984 have been reviewed by Ms.
J Usborne, manager of PTA to verify that no unqualified reviewers reviewed a calc note. No unqualified reviewers were found. September 1 is the date on which both current managers were appointed to PTA/OPA. At this time, PTA and UPA managers are not aware of any cases where an unqualified engineer reviewed a calc note. Thus, there is no impact on plant safety.
Action to Resolve -
As previously committed, the RAT-IG-2 procedure will be revi sed to clarify
, review assignment procedures and ensure management approval of the reviewer. (Osborne, Loftus).
e e
4
, 7. THREATENED RETALIATION FOR SENDING WRITTEN MESSAGES i Dackground.on,the_ Allegation Mr. Seq 1etes alleges harassment by Ms. Osborne Status,_ Safety Impact, and, Action to Resolve Ih t s item alleges no impact on plant safety and has none. No action is required to address safety concerns.
i l
2 l
1
}
O C
1
.I 4
s 4
- 8. POOR CALCULATION NOTE CHECKING WHICH RESULTED IN QUALITY ASSURANCE l VIOLATIONS AND NONCONSERVATIVE COMPUTER INPUT DATA Background on the Allegati.on i Mr. Segletes lists errors he found an a review of CN-TA-84-60. Mr.
Sealetes allegen that the reviewer of the calc note was not qualified.
Status and Sa{ety_ Impact.
CN-TA-84-63 has been completely re-reviewed by Mr. Grace, a qualt+ied revtewer. Mr. J. L. Little has approved the calc note as well. In addition to an overall re-review, the impact of the alleged errors is explained belows
- a. Mr. Segletes states that a model 51 steam generator was simulated' because the variable SGTYPE was set to model 51. However, this LOFTRAN input is only used for steam generator tube bundle uncovery calculations and steam generator water level calculations. Since steam generator tube bundle uncovery is irrelevant in this event and since level calculations are not uned, in incorrect input for SGTYPE has no impact on the results.
The value for MODEPH is correct.
- b. Buoyancy calculations are performed for natural circulation # conditions in the core. Since the transient is over long before natural c i rcul at i'on conditions are reached, the incorrect input of the ZCORE variable will have no effect on the results and conclusions found in the analysis.
- c. This allegation has been addressed in the response to allegation 4.A.
- d. The front page, or cover sheet, of the cale note was completed. See items e, 1, m.
f I
- e. The purpose of the analysis was shown near the front of the calc note, i
~
however, the results were towards the middle. Purpose and results aer,e
- also added to the cover sheet of the calc note. 1
. f. Only two cases were in fact analyzed. The first " case" or computer run was a preliminary run to set up the actual cases analy=ed. The calc note gives a justification as to the cases analyzed and the independent reviewer concurred. The word "four" in the calc note was corrected to read "three".
- g. The table of contents was completed.
- h. The author forgot to label the " Analysis and Methods" section in the calc note so the analysis and methods appeared to be included in the
" Introduction" section. The calc note was corr'ected by adding the
" Analysis Method and Calculations" title on the' appropriate page (4).
i.. Contrary to the allegation, sample calculations are shown in the calc <
note (f or ex ampl e, pages 4, 5, 6, 10, '1 1 , 26, 27, 28, and 29 contain sample calculations). I J. The checklist page was given a page number. Note that the checklist does not add to the input, results, or conclusions of the calculation
- pcrformed.
tho calculation.
If it w2re lost, thcro would bo no information lost concerning I It in uncd only to holp cncuro that tho verification in complete. !
1 k.
NS-RAT-IG-3 states that the cale note must contain " sufficient '
information to be able to reproduce any computer codes used." This does not mean that the listing from every run must be included. This cale note contains major listings and a description on how to make changes to cet to third case. However, clarification of this was added to the calc note, i
1.
The microfiche the calc note.
identification numbers were added to the cover sheet of i
- m. The appropriate manager's signature (Little) was added.
n.
note.
Microfiche identification numbers wera added to the text of the calc This is not required if they are on the cover (item 1).
- o. The information upon which the P-9 uncertainty was based (page 17 of the calc note) did not clarify that the calorimetric uncertainty was j included. The author added an additional 2% uncertainty (see item p).
J This is a conservative results of the analysis.
error and therefore does not adversely aff,ect the -
p.
The 2% uncertainty is not a nuclear flux uncertainty as the allegation statec, but is a calorimetric uncertainty. The text of the calc note was corrected.
q.
The correct value of 4 degrees
- uncertainty was used in the analysis and the text of the cale note which states the total uncertainty value was correct. A correction will be made to the step prior to reaching the total to indicate the uncertainty was 4 degrees not 4%.
r.
Except for the mistake on the LOFTRAN variable SGTYPE (discussed i,n item a), all other inputs corresspond to model D steam generators and are correct. The LOFTRAN input did not assume model 51 steam generators as
' ~ the allegation states since the SGTYPE input variable for this accident is not used. Therefore, there is no inconsistency in the analysis, s.
The input value is correct, however the text should read trip reactivity not shutdown margin. Thus, there is no impact on the analysis.
A correction to the cale note text was made.
t.
Modifications were made corressponding to 60% power which is consistent with the other assumptions made in the analysis. The text in the cale note was corrected. -
. u. The values for the THINC computer code variable NORDER were not changed from the previous analysis from which the computer deck was ob.t ai n ed . The cale note was listing only those c_hances made to the previous computer deck so that it may be used in this analysis. A change was not made to this variable, therefore it did not have to be identified in the text of the cale note. However, the input used for this variable-cculd have been found in the input listings section. There is no error resulting from this item.
)
1 i I l
,...__..____-_____..m. . . _ _ . _ . , _ _ , __ , , - _ . _ - v ,w. , m m .n . .
e
- v. The LUFIRAN variable OFINIL defaults to 1.0, therefore an input for this variable is not really required but is made as a matter of convenience for one who may read the calc note later. Since the correct value was used in the analysis, there is no impact. j
- w. ihe text in the calc note is correct.
- x. Revision 1 of the calc note, made on the same day tho cale note was given final signoff (5-15-84), had already picked up this mistake.
Earlier page by page review of the input and results were done on 4-17-84.
- y. Same as item x above.
- . Page 86 of the calc note contains a letter i dentifying the typographical corrections made in revision 1 of the calc note (also stems x and y). The corrections were clearly identified.
Summary: All of the allegations have been reviewed with the conclusion that noe of.them have any impact on the conclusions of the analysis and there is no safety impact on the plant. Where appropriate, clarification that the " errors" have no impact was provided in t h e r e-r e v i ew'." ' ~
The P-9 analysis is not a standard analysis but is a combination of two standard analyses performed in TA, the Loss of Load and the Loss of Flow.
At the time Ms. Osborne reviewed the calc note, she had performed both of these analyses and was therefore qualified to review this cale note based on the policies in effect at that time (see also allegation 6 B).
Acti on_ to. Repol,ye, None required.
t P
l F
, , , . . . .