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UNITED STATES
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NUCLEAR REGULATORY COMMISSION I
REGION i 475 ALLrNDALE ROAD
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KING OF PRUSSIA. PENNSYLVANIA 19408 APR 0 2199; i
l This letter refers to the following allegations that you provided to us: January 11, 1991, j
alleging that steam generator manway studs were taken from the steam generator mockup and that this was in part, prohibited by administrative controls; January 11, 1991, alleging that the RPS block does not function properly and no corrective actions have been taken; January l
'1,1991, alleging that work order M2-90-15362 was authorized and worked without a
,out: January 31,1991, alleging that Unit 2 is unprepared for implementation of some Appendix R contingency actions; and, February 6,1991, alleging that the containment I
radiation monitor may be isolated because chemistry sampling ficm the monitor does not agree with a grab sample.
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We have initiated action to have the Northeast Utilities staff review your concerns and
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respond to us. We willinform you of their review tmdings. Attached is a copy of the l
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concerns as we characterized them to Northeast Utilities.
f Should you have any further questions, or if I can be of further assistance in these regards, i
please call me collect at (215) 337-5183.
I Sincerely, l
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Gene Kelly, Chi.
Reactor Projects Section 4A
Enclosure:
As stated
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0 ENCLOSURE i
I Issue 1: Some of the steam generator _ manway studs are from the training mock-up. There were no quality controls in place at the mock-up and the studs are not suitable to use on the steam generator manways. A non-conformance report (NCR) was written concerning use of the studs, but the work order was closed prior to disposition of the NCR. The sequence of l
events is prohibited by administrative control procedures.
l Please discuss the validity of the above assertions. If an NCR sas written, pie.ae provide a copy of the NCR for our review. If any procedural requirements were violated, please discuss your corrective actions and the significance to safety of the violation (s).
Issue 2: During a recent performance of SP-2402P, it was identified that the bistable for low steam generator pressure on the Reactor Protection Block does not function properly. A
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modification is underway to correct c:;ft adjustment. A number of other problems wi:h the bistable have been identitled but no corrective actions have been taken.
Please discuss the validity of the above assertions. If problems with an RPS bistable have
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been identitled. p! ease discuss the operability of that channel of :he RPS. Please discuss uhy i
idemified discrepant conditions, if any. associated uith the RPS have not been addressed by management.
t issue 3: Work order M2-90-1530, used to upgrade the flow indica: ion of the containment radiation monitor, RM-8123 was authorized and worked wi:hou: a :agest.
P ease discuss the3ahdity of the above assertion. Was there an administra::.e recuirement to
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have a tago.:: in place to perform the work and why was this requirement not followed?
l P: ease discuss any ger.eric implications.
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ENCLOStRE r
Issue; There are no backup flow indicator instruments for the 10CFR50 Appendix R storage locker. When the instruments are removed for calibration (under EN 21199) there are no controls in place to locate or otherwise provide instrumentation, if needed. Additionally, there are no procedures in place to install the instruments if required by AOP 2579AA.
(This issue was provided to Mr. J. Keenan (Millstone Unit 2) on February 1,1991 in a i
telephone conference wi:h Mr. D. Haserkamp (NRC).)
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when the instruments are place to install the flow indicators if needed, including occa>tur.3
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ENCLOSURE Issue; On February 6.1991, Health Physics Technician (*) sampled the containment air by both grab sample and at the inlet to radiation monitor RM-8123A/B. The samples resulted in about 50 cpm at the radiation monitor and 2300 cpm from the grab sample. The radiation monitor may have been isolated from the containment. Recently, the hydrogen monitor was determined to be isolated from the containment as documented in plant incident report, PIR 91-16.
Please discuss the operability of radiation momtor RM-8123A B on February 6,1991. Was there a procedural problem that caused the t.to samples to differ? Please discuss the possibility of a sahe control problem discussed in the PIR and may have caused the sampling prob:em. If any weaknesses N either prNc,! - comp'iance 0 sa5e control are dc*er-- cd.
please discuss acuens that you have taken or will take to correct the problem.
(*) - identity of technician may be obtained from the Senior Resident inspector.
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m ALLEGATIONS AND COMPLAINTS - GENERAL RI 1210.1/1 il- *"
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APPENDIX 3.1 ALLEGATION RECEIPT REPORT Date/ Time Received:
98 Allegation No. 8 \\ ~ 91-A = do/o Name:
Address:
Phone:_
City / State / Zip:
Confidentiality:
Was it requested?
Yes No X<
Was it initially granted?
Yes No Was it finally granted by the allegation panel Yes No Does a confidentiality agreement need to be sent to alleger?
Yes No Has a confidentiality agreement been signed?
Yes
" No Memo documenting why it was granted is attached?
Yes No dd Position /
Title:
e Facility:
M i 05 k< L Docket No.: 56-636, (Allegation Summary (brief description of concern (s):
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O Number of Concerns:
3 Employee Receiving Allegation:
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(first two initials and last name)
Type of Regulated Activity (a) $ Reactor (d) _ Safeguards (b)
Vendor (e) _ Other:
(c) _ Materials (Specify)
Materials License No. (if applicable):
Functional Area (s):.X(a) Operations (e) Emergency Preparedness (b) Construction (f) Onsite Health and Safety (c) Safeguards (g) Offsite Health and Safety l
(d) Transportation (h) Other:
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(NRC Region I Form 207 Revised 10/89)
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(203) 665-5000 April 26, 1991 Docket No. 50-336 A09352 l
Mr. Charles V. Behl, Director Division of Reactor Projects U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, Pennsylvania 19406
Dear Mr. Behl:
i Millstone Nuclear Power Station, Unit No. 2 RI-90-A-221 We have completed our review of an allegation concerning activities at i
Millstone Unit No.
2 (RI-90-A-221).
As requested in your transmittal letter, our response does not contain any personal privacy, proprietary, or-l safeguards information. The material contained in this response may be released to the public and placed in the NRC Public Document Room at your
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discretion.
The NRC letter and our response have received controlled and i
limited distribution on a "need-to-know" basis during the preparation of this response.
Based upon our request on March 15, 1991, Region I personnel extended the due date for this response to April 26, 1991.
Additional time was requested to support the ongoing INPO evaluation and to i
prepare for an Enforcement Conference held on March 27, 1991 involving an ongoing allegation-related matter.
Issues The following drawing deficiencies associated with the electro-hydraulic control (EHC) system have been identified:
l 1.
A recent modification replaced pressure transmitter PT-4297A and added a current / voltage converter. These items were not reflected on drawing NUSCO 25203-39077, sheet 73D, although PDCR 2-88-83 did update loop diagram 25203-28500, sheet 504.
2.
Drawing No. NUSCO 25203-39077, sheet 73E for the main turbine pressure demodulator identifies a wrong part number (11D9988GE1
[ identified]
i versus 117D9988GE3 [ actual]).
3.
The diode function control board which processes signals to the turbine control is not reflected on NUSCO drawing 25203-39077, sheet 73E.
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Hehl, Director i
U.S. Nuclear Regulatory Commission i
A09352/Page 2
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. April 26, 1991 Please discuss if there is any validity to the above issues and provide the details of your review.
Please include any corrective actions that you have taken or may take in response to the identified deficiencies.
Please j
discuss if any generic draving control problems are apparent as a result of
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your review.
Background
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The referenced drawings are vendor-supplied block diagram schematics showing the configuration of the sy: tem as supplied and are not intended to be detailed viring diagrams. The information missing from the referenced drawings has no effect on the operation of the system, nor does it affect the ability to maintain the system. Other drawings do contain the missing information and would more likely be used to work on the system. These issues could have been easily handled by processing a drawing change request (DCR). Procedure ACP-0A-3.24 provides instructions to anyone who finds a draving in need of change or revision on how to process the proper correction.
Response
Issue 1
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The referenced issues are not considered to be drawing deficiencies. The EHC system is a package system supplied by GE vith these drawings to familiarize the owner with the system configuration.
The missing i
information is not typically supplied on this type of drawing.
- However, the information is contained in other NUSCO controlled drawings. Technical Training is avare that GE drawings do not have a high level of detail and therefore they provide additional information during training on EHC in order to help prevent any confusion. Students are also told where to find the appropriate drawings and information.
A DCR, M2-S-41891, has been initiated in accordance with ACP-0A-3.24 against imSCO drawing 25203-28500, sheet 504 in order to provide reference to draving 25203-39077, sheet 73D.
i This should eliminate any potential confusion in the future.
Issue 2 l
NUSCO drawing 25203-39077, sheet 73E has no part numbers identified on it.
Ve are not able to evaluate this issue with the information available to us at this time.
Please advise us if you vish us to pursue this matter j
further.
Issue 3 The diode function generator control board is shown on NUSCO drawing 25203-39077, sheet 73E, contrary to what is stated. It is located at coordinates A9 of that sheet. While what is represented on the drawing does not
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exactly duplicate the circuit board viring, it is not intended to do so.
As stated before, the drawing is a diagram intended to give a general i
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Mr. Charles V..Hehl, Director e
U.S. Nuclear Regulatory Commission A09352/Page 3
. April 26, 1991 representation of the system configuration. Circuit card inputs, outputs, test points, and points of adjustment are accurately depicted on the diagram.
The details of the make-up of the circuit card are not shown in order to improve the drawing clarity. These details are provided on another vendor drawing (GE 996D336).
After our review and evaluation, we find that none of these issues taken either singularly or collectively present any indication of a compromise of nuclear safety. Ve appreciate the opportunity to respond and explain the basis for our actions.
Please contact my staff if there are any further questions on any of these matters.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY
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_E. J. K oczka f
senior Vice President V. J. Raymond, Senior Resident Inspector, Millstone Unit Nos. 1, 2, cc:
and 3 E. C. Venzinger, Chief, Projects Branch No. 4, Region I j
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