ML20198M963
| ML20198M963 | |
| Person / Time | |
|---|---|
| Site: | Indian Point, Point Beach, Turkey Point, Ginna, Diablo Canyon, Comanche Peak, 05000000 |
| Issue date: | 06/17/1985 |
| From: | Segletes J AFFILIATION NOT ASSIGNED |
| To: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| Shared Package | |
| ML20198M785 | List:
|
| References | |
| FOIA-85-654 NUDOCS 8606050458 | |
| Download: ML20198M963 (21) | |
Text
.
- y. c;- A-d 9201 Wedgewood Drive Pittsburgh, Pennsylvania 15239 June 17, 1985 s
I To:
- Director, Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Sir:
Until July 31st of this year, I will hold the. position' of Senior Engineer in the Westinghouse Water Reactor Divisions / Nuclear Technology Division / Nuclear Safety Department / Risk Assessment Technology section.
Since joining Westinghouse in 1980, I have been located at the Monroeville
~ Nuclear
- Center, Monroeville, Pennsylvania, where I primarily perform accident analyses.
have knowledge of incidents that IOver the past year, I have either bed
~
in or believe are violations of either nuclear regulatory law or Westinghouse quality assurance requirements.
These incidents have been categorized and are shown as Items 1
through 8
on the g
attached sheets.
It is requested that you investigate and take appropriate action where necessary.hese incidences t
If I can be of further assistance, please call me.
.r Sincerely, a..
vohn A. Segletes j
Phone:
(412)795-2795
{
rI 8606050458 860527 PDR FOIA DOHERTY85-654 PDR
1.
LOST DIABLO CANYON SAFETY ANALYSES During a group meeting held in late 1984 for the sPlant Transient Analysis and Operating Plant Analysis
- groups, it was stated by the manager of Operating Plant Analysis 3 Pat Loftus, that almost all of the calculation notes' that support the Diablo Canyon Final Safety Analysis Report are missing.
Apparently these supporting analyses were lost when they were to be put on tape in 1974 To the best of my knowledge, these lost analyses have never been retrieved and no attempt has been made to inform the Westinghouse Water Reactor Divisions Safety Review
, Committee, the NRC, or the customer of this situation.
Also present at the meeting were the Manager of Plant
~
- Analysis, Melita
- Osborne, and approximately twelve engineers and technicians from both groups.
I believe this is a safety violation since not. keeping records that are required by a
licensed condition is a
violation of 10CFR50.71, Part C.
e O-
y/
2 NONDISCLOSURE OF AN UNSAFE PLANT CONDITION AND RETALIATION BY MANAGER During November of 1984, I was assigned the task of l
performing an analysis to evaluate the impact of the flux rate signal device from the Indian Point 2 removing 1
fuclear power plant.
This device is used to initiate turbine runback to protect agai~nst departure from nucleate boiling in case a dropped rod or dropped bank accident occurs.
Redundant protection is provided by a rod-on-bottom signal s
device which also causes a
turbine runback.
The rod-on-bottom device operates concurrently with the flux rate signal device to provide the redundant protection.
Before I started the Indian Point 2 task, I reviewed similar study that was done for the Turkey Point units a
(see,.'
CN-TA-82-104).
It immediately became apparent to 'me that deleting the flux rate signal device at Turkey Point violated the single failure criteria as specified in IEEE 279-1971
- Criteria for Protection Systems for Nuclear i
Power i
Generating St'tions".
This is because the rod-on-bottom
- device, by
- itself, is not totally redundant.
When I
informed my manager (and author of the Turkey Point analysis), Melita
- Osborne, of this violation, she said
" John, do not disclose this information or we will be sued."
(I presume Melita meant Westinghouse would by Florida Power and be sued Light (FP&L)).
After some further discussion on this matter, I dropped the issue because I
believed Melita would retaliate against me if I further.
pursued it Independent of my finding, FP&L later recognized the same unsafe condition existed that I
called to Melita's attention in November 1984 In early 1985 FP&L issued an i
LER to report this problem.
This time Melita did not attempt to conceal the problen nor did she inform FP&L I had previously determined this problem to exist.
The that Westinghouse response to the FP&L finding was documented Letter NS-RAT-PTA-85-091 which provides in recommendations on how FP&L should modify the existing hardware to make the 1
system redundant.
1 On the 29th of January
- 1985, I
had my performance appraisal for the year 1984 and was informed by Melita Osborne that I was being terminated from Westinghouse on July 31, 1985 I believe a factor in my termination was retaliation against me for uncovering this faulty Westinghouse recommendation of which Melita was the originator.
t 4
l
- -__ - _- -. - _.- -- - ~,, -,-,..
3.
FEAR OF RETALIATION
~s As noted in Item 2 shown on the previous page, TI was required to perform a safety evaluation for Indian Point 2,
similar to.the one that Melita had performed for the Turkey Point Units.
In the Indian Point 2 analysis, I stated that the rod-on-bottom unit by itself was not i
completely single failure proof. (see Page 10 of CN-TA-84-202).
On the other hand, I did not disclose this fault in the customer report (NS-RAT-PTA-84-171) since disclosing it would result in either the Indian Point 2 and Turkey Point units having to undergo substantial modifications (to make the rod-on-bottom signal device single failure proof) or t h a '. '
flux rate signal device could not be removed from
- service, which would negate the need for the i
analysis.
based on Melita's response to my finding in the case ofFurthermore,~
the Turkey Point Units, I feared retaliation by Melita if I
disclosed this fault to Consolidation Edison of New York J
City.
I discussed this dilemma with two of-my colleagues, Thomas Blackburn
- and, to, a lesser
- extent, Mark Adler.
Thomas Blackburn later checked my cale note.
e e-l
_.,u..
i 4
NOT REPORTING APPARENT SAFETY VIOLATIONS TO THE SAPETY REVIEW COMMITTEE A.
Analyses That Raise the Reactor Trip on Turbine Trip Setpoint s
In January 1985 a colleague, Mark Grace, pointed out to that me a
Comanche Peak plant specific study he had previously checked (CN-TA-84-97) to justify raising the setpoint for deletion of reactor trip on turbine trip above its then existing value was in error.
The error was the result of not transferring transient inlet temperatures to the THINC3 computer code where they can be used to calculate the departure from nucleate boiling ratio (DNBR).
As a result, the initial (constant) inlet temperature is used to compute DNBR throughout the transient.
This is
.~
unconservative since in some cases analyzed the transient inlet temperature rises approximately 20 to 30 degrees *'
Fahrenheit above the initial temperature by the time the minimum DNBR is reached.
I reviewed our files and determined that' approximately a dozen studies of this type had been done previously an,d in only one of these studies (CN-RPA-78-66) did THINC3 use the correct inlet temperature history.
I wrote a memo on January 25, 1985 to my
- manager, Melita
- Osborne, informing her of the nonconservative computational method currently being
- used, while pointing out that this error probably exists in several other h
studies and that the problem should be reported to the Westinghouse Water Reactor Divisions (WRD)
Safety Review l
Committee (SRC).
When I later spoke to Melita regarding the note, she criticized me for calling the problem to her attention and said she would take care of it.
No plan for resolution of this problem was quickly set-up as required in Risk Assessment Technology (RAT)
- ~
procedure NS-RAT-IG-9 nor was the WRD SRC alerted of this potential issue within the first two weeks as required by Item 10 of NS-RAT-IG-9. To the best of my knowledge, this problem has never been reported to the SRC and it been corrected in two cases.
has only B.
Dropped Rod Analyses for Turbine-Runback Plants In early 1985, Glen Hebele, while working 'on to justify an increase in the turbine runback a
study setpoint for Turkey Point Units 3 and 4 (see CH-TA-95-6),
discovered an error to exist in the dropped rod methodology as outlined
4.(Continued) in NS-TA-83-365 which yielded no~nconservative results!
Specifically the. dropp.ed rod methodology calls for i
performing the analysis at a
turbine runback safety l
analysis limit 4% less than the turbine runback setpoint.
However, the safety analysis limit should be 44 more than the setpoint value.
This 8%
error was incorporated into-the following plant specific safety analyses.
Point Beach 1 and 2 Turkey Point 3 and 4 Indian Point 2 and 3 Ginna Beznau
~
Glen Heberlesinformed me he reported this error'to Melita Osborne, but, no plan of resolution of this problem was quickly set-up in accordance with RAT procedure NS-RAT-IG-9 nor was the WRD SRC alerted of this potential issue within the first two weeks as required by Item 10 of NS-RAT-IG-9.
To the best of my knowledge, the problem was never reported Eo the SRC nor have any of the erronious analyses been corrected.
~
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. -.. - - - - = 5'-
- - - ~ ' - ' ~ ~ - - ~ ~ ~ ~ ' ^
~
~
i 5
NONUNIFORM PROCEDURES WITHIN WESTINGHOUSE FOR REPORTING POTENTIAL SAFETY VIOLATIONS 4
Information provided by Westinghouse management shows three different sets of procedures that should be followed in reporting potential safety violations
,tg the Westinghouse Water Reactor Divlsions (WRD)
Safety Review Committee (SRC).
5 In my opinion, the procedures that are specified by first and second level managers (see B and C
below)
- can, and do, lead to intimidation as discussed in Item 2
and
- burying" potential safety problems as discussed in Item 4 Also, I think it is the intent of the NRC that
- one, and only one, set of procedures be used within Westinghouse to report potential safety violations.
A.
Posted in the main lobby of the Monroeville nuclear Center, Monroeville, Penna.
3 -
(1.)
Report violation to supervisor, or-(2.)
Report violation to Manager's Representative on the WRD SRC, or-
~,.
(3.)
Report violation to R.A.
- Wesemann, Secretary of the WRD SRC.
B.
Stated in the Radiological Assessment Technology Instruction Guidance Material (1.)
Report violation to supervisor, then (2.)
Get supervisors approval, then (3.)
Provide plan for. resolution of the problem to
- the, SRC.
If the supervisor disapproves your request to report potential safety item, you may the
~
(4.)
Report directly to the WRD SRC.
r o
C.
Stated in undated memo provided to members of Plant Transient Analysis and Operating Plant Analysis during a
meeting in late 1984 and also provided in a Nuclear Safety Department handout to all members of the : Nuclear Safety Department in early 1985 (1.)
Get supervisors approval.
Note:
in letter NS-RAT-PTA-85-047I have asked for clarification regarding this issue The response to my request (see Letter NS-RAT-pTA-85-051) states the memo in Item C above was only intended to be a referred to
" guideline",
but there is nothing on the menc to indicate it was only intended to be a guideline.
,,,,_,,,.,~---.;:;-A
- ~ ~ ~ ' ' ~ ~ '
^ '~ ~~
6 VIOLATION OF QUALITY ASSURANCE PROCEDURES A.
Transmittal of Preliminary Draft Reports Letter NS RAT-83-036 dated Noveber 29,1983 stat 5s the requirements to be followed within the Risk Assesment j
Technology (RAT) syction with regards to transmitting i
preliminary draft reports outside the RAT group.
The requirements are the followings (1)
The transmittal letter should state the information is preliminary.
i (2)
The report should be stamped ' PRELIMINARY".
)
(3)
First level manager's approval is required.
? -
j On December 16,
- 1983, a
preliminary copy of the Italian Reference Plant functional requirements were sent i
out (see Letter NS-TA-33-520) without any of the above i
requirements implemented.
Note that these functional l
.=-
requirements did not go through the normal in-house review but rather were to be reviewed by the customer (NIRA/SOPREN).
c.--.
B.
Assigning a Competant Independent Verifier Within the Risk Assessment Technology (RAT) Section 1
Westinghouse Nuclear. Technology Division procedure NTD-DPP-3B, Rev. 2 dated 7/24/81 and RAT section procedure NS-RAT-IG-2 state that the cognizant (or Appropriate RAT) manager shall assig'a an engineer to act as the independent (or RAT independent) reviewer.
This procedure is rarely if ever followed in the Plant Transient Analysis
'or Operating Plant Analysis groups.
In
- fact, I
requested that my
~
manager, Melita Osborne, assign an independent checker to check one of my calculations (CN-TA-85-29) when I
had difficulty in finding an independent reviewer.
Melita returned the calc note later with an attached note stating that I
should find my own independent. reviewer.
Tom Blackburn did check CN-TA-85-29 when I asked him to do so and recalled seeing the note that Melita wrote when I
called the incident to his attention on March 25, 1985 O
7.
THREATENED RETALIATION FOR SENDING WRITTEN MESSAGES On Thursday morning, February 14,
- 1985, my manager, Melita Osborne, called me into her office and told-ae she would terminate my employment with Westinghouse with two months notice if I
continued to harass her.
What she considered harassment included only the following itehs.
1.
Writing Letter NS-RAT-PTA-85-047 which requested clarification on the correct procedure to use to report potential safety problems to the Westinghouse Water Reactor Divisions Safety Review Committee.
2.
An informal meno dated 2/7/85 from me to Melita asking why the normal in-house review and comment procedure was not followed for the Italian Reference Plant'g, -
Functional Requirements (Letter NS-TA-83-520).
"~
3.
An informal meno from se to Melita stating that I
planned to give the Italian Reference plant's Back-up Protection System Functional kequirements c
PRELIMINARY status until they were checked by the customer since this would conform with'NS-RAT-83-036.
4.
An informal memo from me to E.P."' Rahe and D.C.
Richardson regarding a complaint by Consolidated Edison of New York City that Westinghouse had never called back when they (Consolidated Edison) requested a
meeting between 2
Consolidated Edison and Westinghouse a
month earlier.
I also expressed my concern that our good business relationship with consolidated Edison was being strained because of this incident.
Melita demanded that any feature communication I
have with her be limited to verbal communications.
e~
\\
, (
~
8 POOR CALCULATION NOTE CHECKING WHICH RESULTED IN QUALITY ASSURANCE VIOLATIONS AND NONCONSERVATIVE COMPUTER INPUT DATA A review was made of CN-TA-84-63, "CGE Deletion of Reactor Trip on Turbine Trip Below 50% Power (P-9)* 'by E.
l Kurt Hackman and checked by M.P. Osborne.
Numerous _ errors were found to exist in the analysis. Most, if not
- all, of these errors should have been detected by the independent reviewer.
The following errors were noted:
a.
The Model 51 steam generator is simulated in this study.
It does not have a preheater, but the input data (MODEPH-1) indicates a preheater exists.(One or the other of these inputs is in error.)
b.
The buoyancy calculations were to be turned off for conservatism (ECORE=IRVO=2SGT=ESGP=0) per page 9, but ZOORE was in fact set to 120.0.
(This error nonconservative direction.)
is in the The transient vessel inlet temperature, as c.
computed by the LOFTRAN code, increases with time but.this data never
~
got into the THINC3 calculation of departure ~ from nucleate boiling ratio.
(This error is in the nonconservative direction.)
d.
The front page of CN-TA-84-63 is not completely filled out.
(Violates NS-RAT-IG-3 procedure.)
The checklist shows CN-TA-84-63 to contain a
purpose e.
and results near the front.
In fact, a purpose and resulta j
are not shown near the front of the cale note.
f.
The Introduction section (page 3) states four cases were analyzed, but only three cases are shown.
L The Table of Contents on Page 2 is not completed.
g.
- h.. Information that should appear in the
" Analysis Method and Calculations' or " Input Listing
- sections' (pages j
2 to
- 40) are actually put into the Introduction section.
i '.
No sample calculation is shown, but the checklist shows the calc note to contain one, j.
The checklist page is not numbered nor is the cale_ note number shown on the checklist page.
(If this page were separated from the cale note, there would be no way to identify the cale note it came from.)
i
8.(Continued) i s
k.
The input listing for the third case is not shown.
(violates NS-RAT-IG-3 procedure) i 1
The microfiche identification numbers are not on the cover sheet.
(violates NS-RAT-IG-3 procedure) m.
The cover sheet requires a
managers signature, bdt i
there is none, n.
Microfiche identification numbers are not shown anywhere in the calc note.
(violates NS-RAT-3 procedure) s o.
The P-9 uncertainty already includes a
nuclear' f' lux uncertainty.
It is not necessary to account for this uncertainty twice as is done in this analysis.
1
~~
3 p.
The 2% uncertanty noted on page 47 is a
nuclear flux uncertainty, not a LOFTRAN uncertainty.
t q.
On page 11, 5 lines from the bottom /~the last term should be 4
degrees Fahrenheit uncertainty, not 4%
i uncertainties.
I r.
Use of GEND3 indicates a
Model D3 steam generator
~
1 should be used.
The LOFTRAN input assumed the Model 51 steam generator. (One of the two calculations is in error.)
On page 16, DKSCRA=
.04 is not shutdown margin, it is s.
a trip reactivity.
t.
On pages 28 and 29, the statement is made that modifications were made for 52%
- power, but they were in fact made for 60% power, r
o On page 28, no numerical value is given for NORDER.
u.
v.
QFINTL requires an input for each loop.~
The proper input should be QFINTL=3*1.0, not QFINTL=1.0.
w '.
On page 56, third paragraphs *--
a rapid increase in coolant temperature" probably was intended to be a
rapid increase in coolant pressure" x.
On page 62, middle of aecond paragraphs the ipower operated relief valves are actuated, not the safety valves, t
j
. ~
+ -. -....
8.(Continued)
~,.
y.
On page 63, last paragraph:
pressure PORVS should be *-- pressurizer PORVS - *.
z.
There is no indication where the two typos referred to in Revision 1 are located.
They should be clearly marked by a bar in the right margin along with the appropriate revision
- number, but I
don't see any such marking.
(violates NS-RAT-IG-3 procedure. )
~~
A review made of analyses which justify raising ther -
~
setpoint for reactor trip on turbine trip above the typical lot power level has shown the independent
- reviewer, Melita
- Osborne, has never previously performed this type of analysis.
Therefore Melita, or her
- manager, should have disqualified her from being the independent reviewer of this calc note.
wpa m
G D
er e
._.__._,.._.____.____,.._,___.g._-...
A Westinghouse Water Reactor
- T*** Sam Electric Corporation DMslons som PtrisDurgftfennsyfvania 15230
?
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:
4~
This is to confirm our telephone conversation of July 11,'1985 in which we informed you that on that date we first became aware of a letter dated June 17, 1985 and sent to your office by Mr. John A. Segletes, a former employee of Westinghouse e~
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 emp believes were violations of either " nuclear'.loyee which he 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 employee has been inhibited from raising safety concerns
~
through the established channels as defined in the company's molicies 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.
Please call me (412-374-4868) if I can be of further assistance a t any time.
Very truly yours,
~h on n
[
- p. Rahe, Jr., Manager Nuclear Safety Department 1
CF nlp I
9:
NONDISCLOSURE OF AN UNSAFE PLANT CONDITION AND RETALIATIO I
2.
Background on tho, Allegation dropped rod event by the turbine runback system provides protection f or a reducing turbine load to a preset power level.
In the standard design, 1 out-of 4 intiated by either a rod-on-botton or the turbine runback is negative flux rate sianal.
The allocation states that the proposed modification (deleting the flux rate signali violates the requirements of IEEE-279, one of which is the single failure criterion, and that the
~~
is not redundant.
proposed change creates a system which The turbine runback plants are an older catequry of plants fo[r which the licensing criteria, but which were licensing bases do not meet current licenses were issued.
reviewed and approved by the NRC at the time the a safety function in that it must be The turbine runbar.k system performs that the cctuated follow'.nq a dropped rod / dropped bank in order to show (DNBR>1imit value) is met.
However, the original the criteria of licensing criterion the turbine runback system does not meet allis not environmentally or seismic design of IEEE-279, is not fully redundant, Current protection systems meet these is not Class 1E.
to be consistent qualified, and The system design changes were developedsystem to meet requir'ements.
(as opposed to requiring the meet
. with the licensing basisThe f act that the modified system also does'not the the above requirements does not violate the licensing basis ofSegletes and current criteria).
This was the context of the discussions between Mr.
all of plant.
Ms. Osborne.
Status and Safety _ Impact FP&L submitted the analyses performed by Westinghouse to k
in an 'SER dated January 6, The NRC approved the modification 1985,. FP&L had recently implemented a hardware change system.
d on' the Unit 3 turbine runback system which deleted the flux rate input an 1983.*
In February (The criqinal csked Westinghouse to review the hardware modifications.
They proposal had only addressed analyses, not hardware changes.)
A summary of the cpecifically asked about the single failure criterion. Westinghouse re i
l Certain that the ald system met the criteria.
t the criteria to the extent the single the system meet the single failure criterion, but portions of Westinghouse the turbine does not meet the criterion.
FPLL uncid failure of implementation could be done.
cuggested a means by which there-instated the flux rate input) at that time.
(i. e.,
their modification review concluded that none They did not file a LER because their internalFP&L plans to implement the revi upcoming refueling outages at Turkey Point 3L4.
The current analysis
!was needed.
is no impact on the safety of l
approved by the NRC remains valid and there the plant.
Action'to Resolve None required.
I cf8 l
~. -. -.
3.
FEAR OF RET ALI AllON tiackcround Mr. Scalctos perf ormed a similar analysis f or Indian Point.2 such as had been performed for FPLL's l'urkey Point units.
When FP&L asJed questions related to the implementation of the modification, Mr. Seq 1etes was told
'by Ms. Osborne to write a similar letter to Consolidated Edison before they implemented their change in order to ensure that the issue would not arise for Indian Point.
, Status and Safety _Impagt, The letter had not been written when Mr. Segletes left Westinghouse.
The action has been reassigned t,o 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.
Acti q,n to,_Res.olye
, the Issue a letter to Consolidated Edison to clarify the implementation of turbine runback modification.
This will be done by 7-31-85.
(Osborne) h(--
4 L aTTem-td 5 -IRAT-PTq. 85-Z36 ATT AG E.D e
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_ _ _. _ ~
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NS.-RAT-PTA-85-23 8 Q
i l
Nuclear Safety Department FROM:
Risk Assessment Technology WIN:
284-4357/4303 DATE:
July 30,1985 l
SJBJECT: IPP Modification to Turbine Rmback System j
To:
J. Gasperini R&D 18 Keywords: IPP Dropped. Rod.
cc:
P. A. Loftus MNC 4-09A Turbine Rmback G. P. Roulett MNC 4-09 Indian Point O. Meewis R&D 18 FI-85-013 J. L. Little MNC 4-17 R. A. Wiesemann HNC 4-01 E. K. Figenbaun HNC 4-01 In 1984, Westinghouse provided to Consolidated Ectinop p. parety evaluation for a modification to the turbine rmback system at Indian Point Unit 2 which concluded that the deletion of the NIS rod drop (or nux rate) signal input to the turbine rmback-was~ acceptable. This evaluation addressed the transient analysis impact of deleting the nux rate siginal on the socident analysis.
However, precautions must be taken in the'har&are implementation of the nux race signal deletion such that diversity is not removed from the turbine rmbeck protection system. ' This is consistent with Qiapter 14 of the PSAR, which states that there are diverse means for obtaining the rmback. Westinghouse recomends that the deletion of the NIS rod drop signal input to the turbine rmbeck be
' implemented in such a manner as to mairtain consistency with' the FSAR description of diversity, whidt accomts for certain sin 61e failures. The-follcWing provides a backgromd discussion and formal recommendation concerning this issue.
The automatic turbine rmbeck feature at Indian Poirt Unit 2 provides protective action in the event of a single or multiple Rod Cluster Control Assembly (RCCA) drop. Detection of single er multiple dropped RCCAs (including dropped RCCA banks) occurs by either a rod-on-bottom position indication signal or by a change in neutron nux as seen by the excore detectors (NIS rod drop signal).
The rod-on-bottom signal provides separate indication fcr each RCCA in the core and one signal is sufficient to initiate the turbine rmback. Also, a change in flux as seen by one of the four excore detectors will cause the turbine lead to be reeced. 'The turbine Iced is reeced to a preset value. At the same time, l
automatic withdrawal of the control rods is prevented by a rod withdrawal block. This scenario is discussed and analyzed in Section 14.1.4 of the Indian Poirt FSAR.
1 l
l
J'. Gasperini 2
NS-RAT-PTA-85-238 of the following:As stated in the FSAR, diverse actuation of the turbine rm 1.
changer by a preset amount.Re&ction of the load reference setpo setpoint at a constant rate for a preset time.This is accomplished by res 2.
Re&ction of the turbine Iced limit to a preset value.
(a relier valve which limits control oil pressure) is re&ced untilThe Iced lim turtsine thermal Iced as sensed by either of two turbine first stage pressure channels is below a preset value.
such that certain single failures could prevent a rmb 4
single or multiple dropped ACCA. event.
De design of the automatic turbine rmbeck is prone to spurious rmba in the initiation of the rmbeck.rmtseks not caused by an RCCA drop) b (i.e.,
Thus, a single failure of an electrical component (e.g. burn. cut of a rod position indicator signal, failure of one fact, cause a turbine rmback when it is not needed.excore detec plant transients and 'results in a significant loss of operability "andThis causes availability.
rmtacks have occurred over the years. Operating history at Indian Point Un
~~'
I rate input to the rmback logic.he majority of the spurious rmbecks i
f To alleviate this problen, Westinghouse performed a safety evaluation in 1984 to show that deletion'of the flux rate i
(NIS rod drop) signal frem the rmback logic is acceptable Section 14.1.4 of the 73AR. considered the effect of the change on the d The evaluation i
j In the dropped rod accident analysis, certain limiting single active failure fcr the transient are considered.
With the change, it is apparent that for a single dropped RCCA, a single failure of a rod bottom signal would result failure of the turbine rmback.
reanalyzed assuning no turbine rmback.Thus, the single dropped RCCA case was
~
The results show that the static RCCA glisali,tnment am4 dent tamds the, dropped rod if__the niant rernmins in I
cLtrok It was slao stateTthat the dropped bank (multiple dropped RCCA) i analysis was not affected since multiple dropped RCCAs generate several rod i
bottom signals, and only one is required to initiate the turbine rmbeck i
).
I NIS rod drop signal would be adieved.ever, the evaluation did not consi
}fUi d
i failures considered in the origital analysis.such that the sing i
fg As noted above, the original g
turbine load reference re&ction or the turbine Iced limit i
{
rod drop signal provides input to both means of turb De NIS NIS rod drop signal disabled, a single failure (e.g.ine rmback.
Thus, with the non-reendant channel that transmits the rod-on-bottom signals to the turbine an open circuit) in the l.
l e' aim.
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NS-RAT-PTA-85-238 i
load limit and turbine governor control system will prohibit a turbine rmback.
Therefore, protection against a single failure for either a dropped bank or a multiple dropped rod event can only be justified if the failure occurs in the rod-on-bottom sensor / transmitter and not in the non-redundant channel.
Westinghouse recomends that the turbine rmback system be modified such that n\\o$
i single failure could prevent a rmback following a multiple dropped RCCA event when the NIS rod drop signal is deleted. /This could be acconplished modifying the system ao that the rod-on-bottom ' protection is redundan (Figure ndication system should be evaluated. (i.e. Are there reabndant power sup addition, the effects of a loss of power supply to the rod po tion (or is a loss of power fail-safe in that it generates a rmback?) With the single failure condition met the 19814 safety evaluation would remain valid.
Please transmit the above information to Consolidated Edison, to the attention of Art Ginsberg.
Contact the mdersigned if there are any questions.
M "Y
G. H. Heberle J. Polavarapu Plant Transient Analysis Plant Transient Analysis M
. Approved: M. P. Osborne, Manager Plant Transient Analysis m
_ _. _ -. _... _. _ _ _ _ _ _.. - _. -.. _ _ - ~ _ _ _,
FIGURE 1 TURBINE LOAD LIMIT AND TURBINE GOVERNOR CONTROL SY Te-i f
i NOT AfDt!NDANT
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o l
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g CF FW4.L Pow &A P1A SEC J
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FIGURE 2 REVISED TURBINE LOAD LIMIT AND TURBINE GOVERNOR CONTROL SYSTEM
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RET)urCDANT
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B. Dropped Rod Analyses for Turbine Runback Plants Background on the Allegation 1he allecation reports an error in the analysis methodology for the dropped rod event for plants usino the turbine runback feature for mitication of the event.
It states that the error was incorporated into several analyses and that the analyses were not corrected when the error was found by Mr. Heberle.
Mr. Heberle's analysis shows that"'the results f
tre non-conservative only for those cases in which a reactoritrip occurs i
via an overtemperature delta-T signal.
The other cases are not impacted.
The analysis procedure was developed in August 1983 (NS-TA-83-365) as a result of a separate PI on turbine runback plants (PI-83-210) which was closed out in August 1984 (NS-RAW-84-580).
9 S t a t.u s,,an d, Sa f e t.y.. I mp a c_t.
At the time the error was discovered (February 198 a review was made of
( offected plants and it was determined that suffi ont margin existed for the plants pending re-analysis of the accident.
The following list (contains the impacted plants (different than t e list in the allegation)
'7 gnd the status of the reanalysis.
For those plants not yet reana1 7: ed.,
ths availabli h of margin has been reconfirmed informa11y A In ETdition_to DNb marcinf there is margin since the una lyzed p1 nt s are h not at the becinnine of the cycle, which is when th tran en is 4
The reanalysis is therefore required prior to a of the
{11mitina
,e at the latest.
Reanalyses will be done prior to or as part of o
t Reload Saf ety Evaluation.___
fTurkeyPoint3&4
' Reanal y:ed;Cb-TA-85-6 (Error discovered here)
Indian. Point 2 Reanalyzed CS-TA-85-65 g,
' 7 Indian Point 3 Reanal yz ed f.h-T A-85-91 Surry 1&2 Analyzed correctly f or uprating CN-TA-85-47 (Prior analysen are Vepco scope)
Foint Beach 1&2 Reanalysis in progress Ginna Not reanalyzed yet San Onofre Not reanalyzed yet Beznau 1 Per E-Mail from WNI-Brussels (attached) none of the cases trip on OTDT.
Therefore, no impact.
Zorita ENUSA scope.
They may or may not use the procedure.
A1: other turbine runback plants have non-Westinghous,e fuel and the 1983 medhodology was not applied since the analysis was outside Westinghouse ccope.
Aq11,on_to Resolve
-Issue letter to SAS holders formally documenting change to procedure by C-9-85.(Osborne)
-Complete reanalysis f or Point Beach by 8-30-85 (Loftus).
-Perf orm reanalysis f or Ginna by 8-30-85 (RSE due 11-85) (Lof tus).
-Perf orm reanalysi s f or San Onof re by 11-18-85 (RSE due 11-85)(Loftus).
-Notify ENUSA of change in methodology (same letter that changes the procedure).
)
n n
c To: CANTINEAU.B (75:Waj102)
Cc: LITTLE. J.L (WST6162)
Cc: OSB08NE.H. P (WST8665)
From: OSBORNE.M. P (WST8665) Posted: Wed 24-July-85 14:01 EDT Sys 49 (13) 6
Subject:
BEINAU DROPPED ROD ANALYSIS - SELLETES LEITER i
IN THE SELLETES LETTER, BE2NAU IS LISTED AS CNE OF THE ILANTS FOR WHICH THE DROPPED ROD ANALYSIS HAS A NON-CDNSERVATIVE TURBINE RUNBACK ERROR. A POSITIVE UNCERTAINTY IS CONSERVATIVE. HOWEVER, THE ERROR ONLY APPLIES FOR DROPPED ROD CASES WHICH TRIP ON OVERTEMPERATURE DELTA-T. OTHERWISE, IT HAS NO IMPACT.
WE CALLED DAN RISIER ON M)NDAY AND HE TOLD US THAT THE BE2NAU ANALYSIS FOR DROPPED ROD HAS THE ERROR BUT NEVER TRIPS 01 OVERTEMPERATURE DELTA-T. IF So, THEN NO REANALYSIS IS NEEDED. PLEASE CONFIRM THS BY E-MAIL AS SOON AS POSSIBLE SO IT CAN BE ING.UDED IN THE PI FILE.
~-
- REDARDS,
~
E.LITA OSBORNE i
To:
OSBORNE.M.P (49:WST8665)
Cc:
CANTINEAU.B (WEU102)
From:
CANTINEAU.B (WEU102) Posted:
Thu 25-July-85 14:50 BST BRITAIN p
Subject:
NOK DROPPED ROD ANALYSIS INT.REF. SA/85/564 INTERNAL COPY: D.RISHER
'FIN REPLY TO YOUR E-MAIL OF 24 JULY REGARDING NOK ROD DROP ANALYSIS I CONFIRM THAT OVERTEMPERATURE DELTA-T REACTOR TRIP DOES NOT OCCUR FOR ANY CASE ANALYZED, EVEN THE CASE OF 0 PCM DROPPED ROD WORTH (IE k SPURIOUS RUNBACK). THIS MEANS THAT THE DIRECTION OFiTHE ASSUMED s
g RUNBACK SETPOINT ERROR HAS NO IMPACT ON THE ANALYSIS.
COULD YOU PLEASE SEND US A COPY OF THE SEGLETES LETTER 7 THANKS.
REGARDS BOB PRIOR SYSTEMS ANALYSIS, BRUSSELS l
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