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'o UNITED ST ATES 83 NUCLEAR REGULATORY COMMISSION M'
E ntcioN :
47s AtttNDALE ROAD i
j (tNG or enusstA. PENNSYLVANIA 1%X>141s MG16591 I am responding to a concerns that you brought to the NRC's attention on May 6,1991, regarding compliance with GDC 17 - Offsite Power and protective relaying coordination for Millstone Unit 2, and possible harassment of technical staff.
We have takeu action to inspect the technical issue of GDC 17 compliance, as documented in NRC Inspection Report 50-336/91-15 (Section 6.4). This issue has been designated as an unresolved item and, additionally, has been referred to our Office of NRR for licensing review.
Our preliminary conclusions are that these concerns represent reliability issues, of relatively low risk, and are primarily a licensing question. Nonetheless, we will follow the technical issue to closure as part of unresolved item 91-15-02. The potential harassment issue has been provided to our Office of Investigations.
As such, we consider our response to your concems to be complete. Should you have any further questions, or if I can be of funher assistance in these regards, please call me collect at (215) 337 5225.
l Sincerely.
(u.-
u Ed ard WeniNger, Chi Reactor Projects Branch 4
Enclosures:
As stated Inin2:i:n in this ts:::d v;as ds!Med in IcMare r;;th the itadom of !nictmaSon Act n Ti.: _ I d 2 C t
f31 M d6 h OFFICIAL RECORD COPY 9303220399 921217
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Allegation File, RI-91-A-0091 E. Conner 1
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W. Raymond E. xelly EG&G Idaho Representative (CARDONE)
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UNG oF PRU$stA. PENNSYLVANIA 154 % 1415 e,
AUG 161991 T
I am responding to a concerns that you brought to the NRC's attention on May 6,1991, regarding compliance with GDC 17 - Offsite Power and protective relaying coo'rdination for Millstone Unit 2, and possible harassment of technical staff.
We have taken action to inspect the technical issue of GDC 17 compliance, as documented in NRC Inspection Report 50-336/91-15 (Section 6,4). nis issue has been designated as'an unresolved item and, additionally, has been referred to our Office of NRR for licensing review.
Our preliminary conclusions are that these concerns represent reliability issues, of relatively low risk, and are primarily a licensing question. Nonetheless, we will follow the technical issue to I
I closure as part of unresolved item 91-15-02. ne potential harassment issue has been provided to our Office of Investigations.
As such, we consider our response to your concems to be complete. Should you have any I
further questions, or if I can be of further t.scistance in these regards, please call me collect at (215) 337-5225.
Sincerely, 1
% 4er, Chi Wendig Reactor Projects Branch 4
Enclosures:
As stated I
bec:
Allegation File, RI A-O}91 E. Conner T. Shediosky W. Raymond E. Kelly EG&G Idaho Representative (CARDONE) l$caban in B.s n: cts m: nnj in 20Zd3000 U.th the frcdr ' gm3;;0n Atl. cum /icr;s _3 f 7 0 FOIA _
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27 The AV solenoids were replaced with qualified solenoids during the forced outage which began on May 25.
Conclusion and Assessment This item represents a lack of thorough engineering evaluation to appropriately classify components. The engineering evaluation was dencient in that the AV failure mechanism was not fully evaluated with respect to the integrated system interaction.
Further, once the results of the reevaluation were known, NNECO engineering did not inform the operations department on a timely basis regarding the potential operability impact on the EDGs. Specifically, it was known by NNECO engineering on approximately May 2, that the solenoid failure mechanism could affect operation of the EDGs yet the operations department was not formally made aware of this issue until May 28. The judgement and conclusion of the operability assessment were adequate.
NNECO questioning of past quality assurance controls for the AVs exemplified a good safety ethic. The untimely communications had no significance in this case since there was no impact on diesel operability. The NRC is concerned that the communication of potentially reportable issues could be improved.
6.4 Electrical System Compliance with GDC 17 NNECO performed a safety system functional inspection (SSFI) on the reactor building closed cooling water (RBCCW) system in August 1988. The inspection reviewed the ability of the RBCCW system to perform its intended design and safety functions and included an evaluation of the adequacy of emergency power for RBCCW components. At Millstone 2, i
two redundant diesel generators provide the Class 1E emergency AC power for the RBCCW and other safety systems. Several deficiencies identified during the SSFI were dispositioned at the time of the inspection and resolved.
The electrical system issue discussed below was SSFI Observation No. 61 and is the last item to be resolved by NNECO. The issue involves the postulated loss of the emergency diesel generator (EDG) supply under cenain scenarios involving a faulted condition combined with the occurrence of a limiting single active failure. NUSCO engineering reviews of the electrical issue raised the question as to whether the Millstone 2 electrical system design was in compliance with certain 10 CFR 50 Appendix A general design criteria (GDC), and specifically with GDC 17, Electric Power System Design. Compliance with GDC 17 is a requirement for Millstone 2 as a condition of the plant operating license.
The issue was discussed within NUSCO, and after different engineering groups could not arrive at a consensus, NNECO hired an outside consultant to study the matter. The results were reported to the Manager of Generation Facilities Licensing by a memorandum dated April 30,1991, with the conclusion that Millstone 2 was in compliance with GDC 17. At 1
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28 the end of this inspection period, the matter was still under NNECO review with the study f
recommendations presented to the NUSCO Vice President for Engineering Services for i
dispositioning.
Millstone 2 AC Electrical System Design f
At Millstone 2, AC electric power is provided by the offsite power system comprising a l
normal station senice transformer (NSST) and a reserve station senice transformer (RSST),
i both powered from the 345 KV switchyard. The NSST is powered by the Millstone 2 main i
generator and provides normal power to 4160 volt non-Class lE buses 24 A and 24B. Class i
IE buses 24C and 24D are powered from the non safety buses via cross tie breakers. Buses 24A and 24C comprise onsite electric division A; buses 24B and 24D comprise electric division B. Refer to the unit two electrical distnbution system figure below.
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.w 2.im:; E- :dau-J 29 On loss of normal power, Class IE buses 24C and 24D, can each be powered by an emergency diesel generator (EDG) in about 10 seconds. Additionally, the RSST provides automatic backup power to buses 24C ano 24D in less than one second. One type of protection relays senses an undervoltage condition on the safety buses, which operates to isolate the safety buses 24C and 24D and to start and connect the EDGs to the buses.
Another type of protective relay senses overcurrent conditions indicative of an electrical fault.
The 4160 volt overcurrent protection scheme is coordinated to isolate the unaffected portions of the AC distribution system from a faulted section. The issue essentially involves the lack of coordination between the overcurrent and bus undervoltage protective relay schemes used in the 4160 volt system.
Postulated Fault Scenarip The GDC 17 issue involves the response of the 4160 volt AC distribution system to the following postulated scenario. The plant must be designed to satisfactorily respond to a postulated large break loss of coolant accident (LOCA) followed by a single failure. The LOCA will result in an automatic reactor and turbine trip and the automatic transfer of both onsite electric divisions to the offsite RC,ST supply. If the single failure is a fault on one of the onsite electric divisions (e.g., Division A), excessive current will flow on the faulted bus.
I_ow voltage will occur on the Division A and B buses, since both are fed from the same transformer winding. The undervoltage relays will trip before the Division A supply breaker on overcurrent to isolate the fault. Thus, instead of the fault being isolated by a trip of the Division A overcurrent relays, the fault affects both divisions and appears to the operators to be a loss of offsite power. Since no overcurrent relays acuvate, no indication is available to the operators to signal that an overcurrent initiated the undervoltage trip condition.
Once divisions A and B have been isolated from the RSST by the operation of the undervoltage protection scheme, both Millstone 2 EDGs will stan and connect to their respective buses. The division A EDG should then trip on overcurrent as a result of the fault remaining on the division A bus. The division B bus would be powered from its EDG, unaffected by the division A fault. The systems and equipment powered by division B alone has the necessary capacity and capability to cool the core and safety shutdown the plant should this scenario occur.
Event though the breaker supplying power to division B from the RSST tripped as a result of the undervoltage relay operation, the breaker could be shut by the operator from the control room to re-energize division B from the offsite supply. This action should not be necessary immediately because of operation of the division B EDG. However, if such a fault occurs during a LOCA, it is doubtful the offsite circuit would be restored by the operator within a few seconds because of the lack of overcurrent indication and due to the activity in the control room. The question raised by NNECO is whether Millstone 2 meets the requirements i
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30 of GDC 17 if the AC distribution system design (inclusive of the protection schemes) would f
not preclude the division A fault from causing the loss of immediate supply of power to the unaffected onsite electric division from the offsite RSST supply.
j GDC 17 Requirements j
NNECO review determined that the following GDC 17 requirements were the most relevant to the question regarding compliance with the offsite circuit criteria. Two electric power l
systems are required, onsite and offsite. Each power system shall provide its safety function I
assuming the other is not functioning. The safety functions are to maintain reactor design limits during anticipated transients and to cool the core during postulated accidents.
l The onsite power system shall meet the single failure criterion. Offsite power shall be j
supplied by two physically independent circuits designed to minimize the likelihood of their j
simultaneous failure. Each of the two offsite circuits shall be available in sufficient time to i
prevent exceeding reactor design limits, given a failure of the onsite power supplies and the other offsite circuit. One of the offsite circuits shall be designed to be available within a few seconds following a LOCA to assure adequate core cooling. Provisions shall be included to '
i minimize the probability of losing electric power form any of the remaining supplies as a j
result of loss of the other supplies.
f The conclusion that the Millstone 2 design is in compliance with the GDC 17 relies on the
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NNECO interpretation that the offsite circuit design need not consider single failure of the j
onsite power system during accident conditions.
t Conclusions NNECO used a probablistic risk assessment to estimate the safety significance of this issue.
I NNECO concluded that the contribution of the current protective relay configuration to the l
risk of core melt was very low at less than IX10-7. Based on the above, the inspector concluded the issue had low safety significance and that no further immediate NNECO acuan r
was warranted pending the completion of NRC management review of the issue.
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l The question of whether Millstone 2 meets the requirements of GDC 17 is a licensing m !ter.
This issue was referred to NRC:NRR on May 20 for review to determine whether further action by NNECO for Millstone 2 is required. This matter is unresolved pending further review by the NRC staff (50-336/91-15-01).
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i 6.5 Steam Generator Tube Inspections Backeround During plant operation at 100% full power on May 25, a primary to secondary leak developed on No. 2 steam generator (SG). The leak rate increased from approximately zero i
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y1 ll1 ALLEGATION RECEIPT REPORT l~Y'd0?0 NAY Q l%i ll45DM.
A11egation No.
ce Name Address:
Phone:
City / State / Zip:
Confidentiality:
Was it requested?
Yes No '
Vas it initially granted?
Yes.
No Was it finally granted by the allegation panel _
Yes No.
l Does a cenfidentiality agreement need to be sent j
to alleger?
Yes No r
Has a confidentia11ty agreement been signed?
Yes No j
Memo cocumenting why it was-granted is attached?
Yes No
$ %,'2 JJUEdo p,3313,nf7,13,,y N
ON 2 Docket No.: 3-33b Facility:
t (Allegation Su5NemS(,brief description of concern (s): h. Mf6 M5 l
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Number of Concerns:
I Eeployee Receiving Allegation:
P J MA6/6HOEST k (first two initials anc last date)
Type of Regulated Activity (a)
Reactor (d) _ Other: Safeguards
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(b) _ Vendor (e) _
(c) _ Materials (Specify)-
Materials License No. (if applicable):
Functional Area (s):
(a) Operations
_(e) Emergency Preparedness (b) Construction (f) Onsite Health and Safety j
(c) Safeguards
_(g) Offsite Health'and Safety (d) Transportation (h) Other:
(NRC Region I Form 207 Revised 10/89)
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