ML20056B653
| ML20056B653 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 10/29/1971 |
| From: | Mcelroy D NORTHERN STATES POWER CO. |
| To: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML20056B654 | List: |
| References | |
| NUDOCS 9102070674 | |
| Download: ML20056B653 (9) | |
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Dr Peter A Morris M
Director Division of Reactor Licensing United States Atomic Energy Commission Washington, DC 20545 MONTICELLO NUCLEAR REACTOR GENERATING PLANI DOCKET 50-263 LICENSE NUMBER DPR-22 Dear Dr Morris Your letter of October 18, 1973 is acknowledged.
- Enclosed, herewith, is a copy of a memo to me from Mr Lee O Mayer, Manager, Power Production, who is directly responsible for operation of the Monticello plant.
Mr Mayer's memo contains a point by point response - and in most cases, a refutation of the technical aspects of issues raised in your letter of October 18.
However, your letter raises in my mind some issues and questions which go beyond the technical response.
I want to address some of them and that is why I am writing this letter to you.
First, let me lodge a strong complaint.
Your letter gives the impression you have had these concerns for some time.
If so, I am unable to understand why they have been held in silence for so long.
I am extremely distressed you wrote such a letter with its very questionable assertions without first extending us the courtesy of discussing your concerns either over the phone or inviting us to discuss it with you over the table.
Second, in reading Mr Mayer's memo, I sense that several portions of your letter concerning " unusual occurrences" and alleged safety implications are based on misunderstanding and erroneous information, and generally evidenced a lack of under-standing of at least this specific reactor.
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I Dr Peter Morris 2
October 29, 1971 There is one possible explanation for this information gap which, though understandable, does little to comfort me.
It seems to me Mr Mayer's letter pinpoints a very large communi-cation gap between your office (Division of Licensing) and the Division of Compliance in Chicago.
As you well know, the regulations contemplate the Division of Compliance as the primary liaison for a plant once it goes into commercial operation.
Communication with the Division of Licensing is therefore more limited than with the Division of Compliance.
For example, Dr Mann and Mr Benaroya to my knowledge are the only members of the Division of Licensing staff who have even vicited the plant since it went into com-mercial operation last July 1.
Even those visits had specific and limited purposes.
Dr Mann simply wanted to become familiar -
with the facility, its operating staff, and their training.
Mr Benaroya's visit was to secure background information (termed l
an " informal" visit) to aid him in evaluating our application for installation of the off-gas hold up facilities.
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of those facilities, the application is now seven months old and a review by your staff is now in the process.
The early
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installation of these facilities is important both to NSP j
l and to the people of this state.
I As further evidence of an apparent communication gap within the AEC, I enclose a letter dated October 19 from Boyce H Grier, the Regional Director, Division of Compliance con-cerning investigation of the July 14 unplanned release.
At-tention is specifically directed to the third paragraph which s ta tes :
"No items of noncompliance with AEC requirements were identified within the scope of these inspections."
1 find it difficult to reconcile portions of your letter with i
Mr Grier's letter.
I recognize that with Mr Price's resignation and the appointment of a new chairman, coupled with the impact of the Calvert Cliffs decision, there will be an understandable period of adjustment and confusion.
However, 1 sincerely hope that your letter does not herald a new approach to AEC relationships.
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If you have any questions, or would like to discuss my great concern and dismay concerning your letter, I would be very pleased to come to Washington at any time that is convenient to you and the members of the Commission to discuss this matter in any depth you or he Commission desire.
Sincerely x
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NORTHERN STATES POWER COMPANY usuu s A*ou s. wou s e sotA es4os 1
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Oh Daar Mr McElroy:
(n MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO 50-263 LICENSE NO DPR-22 The AEC Division of Reactor Licensing (hereinafter referred to as DRL) letter of October 18, 1971 from Dr Peter A Morris in regard to the handling and res-olution of several operating occurrences at the Monticello Nuclear Generating Plant was received on October 22, 1971 and requested that an answer be sub-mitted within 10 days of its receipt.
j We are of the opinion that the Monticello Plant operations have been conducted in a conservative and safe manner with detailed and deliberate review of sig-l nificant operating events by highly knowledgeable personnel on the plant staff, the Operation Committee and members of the Northern States Power Company General Office operating and engineering groups. Monticello activities are reviewed and q
audited by the Safety Audit Committee. It is our opinion that the Monticello Safety Audit Committee possesses a substantial depth of talent in pertinent dis-
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ciplines and is so organized and staffed as to provide a substantial amount of objective analysis independent of our line operating organization.
We believe that pertinent information concerning the four operatin6 occurrences referred to in the DRL letter, and which should support our contention of ade-quate safety review, might not have been fully available to DRL.
It has been i
our understanding that for an operating plant, our regular contacts with the AEC should be with the regional office of the AEC Division of Compliance (here-inafter referred to as Compliance). Oar philosophy has been to report frequently, and with t very conservative interpretation or definition, all significant events at the Mont: cello Plant.
Compliance has, under this policy, received frequent and detailed luformation on the four items referenced in the DRL letter of October 18.
'ihe current letter and a previous letter from DRL dated July 26,1971, both state that because the flow restrictor pressure drop is not as expected, the flow restrictors may not function to the extent assumed in the safety evalus-4 tion.
There is no evidence to indicate that the restrictors will not provide 1
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THERN STATES POWER MPANY Mr D F McElroy October 29, 1971 the expected restriction in flow. Inspection of the physical installation, review of specifications and records, and radiographs of the "C" steam line flow restrictor have established that the restrictors are correctly sized and properly installed. The design of the restrictors has been thoroughly analyzed and found to be structurally sound.
Ihily readings of the flow restrictor pres-i sure differentials are being recorded to detect any deviation from expected per-formance.
Following the DRL verbal request of July 7,1971, and which request is referenced in the DRL letter of July 26, 1971, the setpoints of the high flow sensors were immediately lowered to a point corresponding to a tested flow level (113% of rated L
flow) and have been maintained at this tested level ever since.
DRL also requested in the July 26, 1971 letter that the sensors be tested at the 'D chnical Specifica-tion setpoint of 140% of rated flow, subsequent to submission a: 3 approval of plans and methods to be used to conduct the tests safely. Preparation of a test procedure and securing the necessary safety analysis information by the Monticello Plant staff and evaluating the safety aspects of performing this test by the normal review groups has been continuously pursued.
The program has involved considerable
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correspondence and several trips to the reactor supplier's facilities to assure ourselves that we had adequate infon::stion to support a finding that the test could be safely performed.
The test procedure and the supporting analysis necessary to perform this test are now finalized and have been thoroughly reviewed and approved by the Operations Committee and the Safety Audit Committee. A Technical Specification vaiver nee-escary for conduct of this test is being prepared and will be transmitted with the proposed test procedure and analysis to DRL shortly.
Daring this interim period, we have kept Compliance fully informed on the substance and progress of the safety evaluation and procedure preparation. We specifically discussed with Compliance the results of tests performed by the reactor rnnufacturer which have established the existence, and cause, of internal leakage paths within the flow restrictors.
This condition is responsible for the lower-than-expeeted values from the differential pressure sensors, but should have no effect en the flow restricting capability of these devices. We soon expect to receive a written report from the reactor manufacturer on these tests and we vill make the infor-mation available to DRL as a part of our program to investigate the "anomolous behavior" of the sensors.
Compliance is also well aware that we i= mediately complied with the DRL request to keep the high flow sensor setpoints within previously tested levels.
We have full confidence in the prudence and safety of operating the plant in this manner during preparation of test procedures for 140% flow testing.
With reference to the July 14, 1971 unplanned activity release as covered in the third paragraph of the DRL letter of October 18, 1971, several statements should be revised on the basis of current information. Although the reactor building vent release rate durin reached 20,000 uCi/see,g the July 14 episode was initially reported as having subsequent evaluation conducted jointly by the plant staff and Compliance representatives has shown that the release rate from this vent was 2,000 uCi/sec.
Contrary to the statement in the DRL letter of October 18,
N THERN GTATED POWER MPANY Mr D F McElroy october 29, 1971 our determination of the discharge canal activity on the day of July 14 indicated that the isotopic limits of 10 CFR 20 were not exceeded.
he October 18 letter also states that DRL informed us, in an August 4,1971 dis-cussion on this matter, that the precautions we had taken with aspect to the air ejector system and operating procedures would not prevent similar releases and that a means to prevent the loss of the off-gas loop seal should be provided.
The NSP personnel involved in that discussion cannot recall such a request, nor do we have any written request of this nature from DRL or any other Division of the AEC.
Had we been aware of such a request, we certainly would have provided a response.
It is our opinion that the actions we have taken as a result of the July 14 release vill prevent similar releases in the future. It should also be noted that the proposed off-gas system modifications for Monticello which have been submitted to DRL for review do include changes which will completely elim-inate concern about the loop seal.
For your further information, four individuals representing Compliance made in-spection trips to Monticello on July 20, 1971 and August 3,1971 to review the July 14,1971 unplanned release. The individuals reviewed the immediate actions taken, the corrective action planned and implemented and other pertinent items.
In the Regional Compliance D.irector's report to us dated October 19, 1971, it was stated "No items of noncompliance with AEC reqairements were identified within the scope of these inspections." The report did not request any further off-gas system modification studies or procedural changes. In the absence of any such comments or recommendations, we must conclude that all AEC concerns with respect to this event have been resolved.
In regard to the item on the HPCI system contained in the fourth paragraph of the DRL letter of October 18, 1971, the 'NX of April 28, 1971 was submitted in accordance with sections 3 5.D.4 and 3 5.H of the Technical Specifications.
These sections are reproduced below:
3 5.D.
High Pressure Coolant Injection (HPCI) System 3 5.D.I.
Except as specified in 3 5.D.2 below, the HPCI system shall be operable whenever the reactor I
pressure is greater than 150 psig and irradiated fuel is in the reactor vessel.
3 5. D.2.
From and after the date that the HPCI system is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding seven days unless such system is sooner made operable, provided that during such seven days all of the Automatic Pressure Relief system, the RCIC system, both of the core spray systems, and i:
the LPCI subsystem and containment cooling mode of the RER system are operable.
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O NJTHERN CTATED POWER MPANY Mr D F McElroy october 29, 1971 3 5.D.3 To be considered operable, the HPCI system shall meet the following conditions:
'lhe HPCI shall be capable of delivering 3,000 gpm a.
into the reactor vessel for a reactor pressure range of 1150 psig to 150 psig.
b.
The condensate storage tanks shall contain at i
least 75,000 gallons of condensate water.
I c.
The controls for automatic transfer of the HPCI
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pump suction from condensate storage tank to the j
suppression chamber shall be operable.
h 3 5. D.h.
If the requirements of 3 5.D.1-2 cannot be met, either 3 5.H shall be complied with or procedures for an orderly reactor shutdown shall be initiated immediately and the reactor pressure shall be reduced to 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.
(Note - emphasis added.)
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3 5.H.
Extended Maintenance l
When it is determined that maintenance to restore l i.!
components or systems to an operable condition vill I
last longer than the periods specified, a report ij detailing the circumstances and the estimated date for returning the components or systems to an operable condition shall be submitted to the AEC prior to the allowable end of the out-of-service period.
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At the time, we acted with the understanding that the letter, and the intent, of sections 3 5.D.4 and 3 5.H allowed such time extensions.
In the DRL reply of April 29, 1971, in which they denied our request for several additional days with HPCI out of service, they clarified the intent of the Specifications by i
stating that it was meant to allow continued power production under "certain l
critical conditions". DRL stated that during startup testing, when all systems have not been completely tested, the availability of safety systems is especially essential. We had no disagreement with this position and proceeded at once to comply fully with section 3 5.D.h of the Technical Specifications.
There is apparently an incomplete understanding of the HPCI system isolation interlocks and of the configuration of the HPCI steam line with respect to the main steam isolation valves.
The HPCI steam supply from the reactor is upstream from the "B" steam line isolation valves and is unaffected by closure of a "B" steam line.tsolation valve. We have stated that the HPCI system would be tested I-l.
N THERN OTATES POWER MPANY P.r D F F2cElroy October 29, 1971 future by isolating the "B" steam line because that is the only test con-in tt diti n with the present system under which hTCI rated pressure and flow conditions can.e reliably achieved without risk of a false EPCI isolation. This has not been stated as a " solution"; in fact, it results in considerable operating $n-convenience. The statement in the D?l, letter that this vill require a reset action by the operator within the 45 second time delay to bring the system back into operation is not correct. The statement that our "colution" is not accept-able because " automatic actuation of all engineered safety features is required" is confusing. Testing of the EPCI system with the "B" steam line isolated elim-instes the possibility of the generation of a false EPCI steam line isolation signal. Therefore, reset action by an operator is not required. Also, auto-matic actuation of the system is completely unaffected. With the "B" line iso-lated, reactor steem is still available to the EPCI system and the system vill function as designed.
Testing the system with the "B" steam line isolated in no way causes or creates conditions urrler under which the EPCI system "may isolate when needed".
The cause of the false signals which produced EPCI isolations has been found and is as reported in the ESP letter to DRL dated September 28, 1971. Our evaluation has shown that the system will perform its designed functions when required and safety and reliability have not been compromised.
We plan to investigate practicable methods of improving this system.
We must demonstrate, however, that the means of eliminating the problem does not affect safety and reliability.
j We believe that the following coments vill clarify the situation concerning the
" loss of instrument air" occurrence en September 5, 1971 which is covered in the
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fifth paragraph of the DRL letter of October 18, 1971. The statement that the
.eactor was restarted before investigating the causes of the loss of air pressure occurrence is not correct.
Investigation was made by plant supervisory personnel i=ediately following the event.
Such investigation is always made following scrans or other significant operating events. The main objectives of these investigations are to identify-1.
The cause of the scram or disturbance 2.
Any equipment failu es er malfunctions 3
ArJ procedural errors er inadequacies l
These are then evaluated and appropriate corrective actions are then prior to startup.
The immediate investigation on September 5 included discussions by technical staff representatives with the operators and the Shift Supervisor on duty during this event, review of pertinent recorder charts, and review of the computer logs.
In evaluating the available infomation during this immediate review, it was con-cluded that a Group I (main steam isolation valves, main steam drain isolation valves, reactor recirculation system cample isolation valves) isolation had not j
been initiated.
Subsequently, as part of the review of the occurrence for the purpose of preparing a written report, a detailed analysis of computer logs and steam pressure recorder traces indicated that the main steam isolation valves had closed and then reopened a short time later.
Since none of the operators was aware that a Group I isolation had occurred, and since they could not recall having
N THERN STATES POWER MPANY Mr D F McElroy October 29, 1971 reset the isolation circuitry, we immediate3y became concerned that the isolation seal-in might be inoperative. At this time it was also considered possible that the valves had closed and reopened as a result of the reduction and recovery of air pressure which had initiated the occurrence. The plant staff directed that the plant be shut down to enable testing of the MSIV's and Gmup I isolation 1
sensors and logic. The tests definitely established that the reduced air pres-sure was not responsible. Complete testing of the MSIV's, the isolation instru-mentation, and the logic circuits, including the seal-in circuits, showed every-I thing operated as designed. Since it vould require a double failure of active components, or, at least a double short-circuit to cause a seal-in failure of all j
eight MSIV's, the probability that an intermittent failure had occurred was very low. The Group I isolation reset switch is identical to and located adjacent to l
the Group III (reactor cleanup isolation valves) isolation reset switch. The j
Group III isolation must be, and is, routinely reset following every scram from high power. Because of repeated Group III isolation due to high temperature at the inlet to the cleanup demineralizers, it is common to have to reset more than It is possible that the operator accidentally reset Group I vhile attempt-I once.
ing to reset Group III. The Operations Committee concluded that this was what must have occurred, a conclusion reached only after inspections and tests had eliminated every other known possibility. We believe that the depth of the sub-sequent investigation and the concern and thoroughness in pursuing a possible safety system malfunction were demonstrated by the determination that the valves had isolated, and by the decision to-shut navn the plant to perform complete tests of the system.
4 Significant operating events, such as those referenced in the October 18, 1971 DRL letter, result in prompt attention by several groups as soon as they occur
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or are recognized. The plant staff takes immediate action to assure safety of j
people and equipment. They also keep the NSP General Office technical staff and management regularly advised on plant and safety status as a result of the j
event and solicit the help of specific talents from that group as needed; they j
keep the Compliance group posted as appropriate; they solicit _ advice from Safety Audit Committee members and independent consultants as required; and, of course, prepare the necessary reports and documentation of these occurrences.
I We have reviewed the handling of the various events referenced in the October i
18, 1971 DRL letter and are of the opinion that safety was not compromised and i
that appropriate management and committee review attention was provided. The
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Safety Audit Committee has reviewed each of the referenced occurrences in detail i
and has not identified any unresolved safety questions. These reviews vill be documented in the usual manner in the Operation Committee and Safety Audit i
Committee minutes and the minutes are available for review by Compliance rep-
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resentatives during their site visits.
s If DRL, after reviewing the additional infomation presented herein, supple-
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mented by verbal inquiries which they might wish to make for further technical i
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N THERN STATES POWER MPANY Mr D F McElroy October 29, 1971 clarification, still have concerns or questions regarding our nnnngement and organizational structure and capabilities, ve believe that these concerns can be most effectively pursued by having appropriate representatives from our General Office staff discuss this subject with DRL representatives at their offices. We stand ready at all times, to satisfy AEC concerns in person at their offices, at our headquarters, or at the plant.
5 Yours very truly, h.
1, L 0 Mayer, P.E.
Manager of Power Production lOM/ma