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NUCLEAR REGULATORY COMMISSION 3
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KING OF PRUSSIA, PENNSYLVANIA 1945 1415
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==Dear
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l In our letter to you dated July 29,1991, which responded to your concerns that you brought to t
our attention on May 8,1991, we stated that we would inspect the incident associated with your corcerns and would inform you of our findings.
We have completed the inspection of your concern regarding the work order used for handling of a Source Range Monitor (SRM) dry tube on May 1,1991, at Millstone Unit 1. Our findings are documented in Millstone Unit 1 Inspection Report 91-06.
Also, as discussed with Gene Kelly of my staff on August 27 but not referred to in our August 14, 1991 letter to you, we have compared the May 1,1991 Unit 1 event with a previous Unit 2 event that occurred in February 1989 as documented in Unit 2 Inspection 89-05. We acknowledge your confusion as to how our August 14 letter did not address similarities with the more recent Unit 1 event since your concems were relayed to us at different times (May 6 versus May 8) and we therefore have addressed them separately. As discussed verbally on August 27 with Mr. Kelly, while there are gross similarities with respect to handling of an irradiated object and movement through the refueling canal, the administrative controls (including procedural adequacy), work practices and radiological effects in each instance were markedly different. No funher NRC action is therefore planned, and we consider this issue closed.
Our assessment is that the issues in question for incore instrumentation at Unit 2 in 1989 which were panly substantiated involved the adequacy of pre-job briefings, use of a hoist / winch, presence of a senior licensed individual, exposure controls and health physics precautions, and management direction. However, no actual incident occurred at that time. More recently on Unit 1, the SRM dry tube did in fact momentarily break the pool water surface, but all of the above aspects were found to be acceptable. Further, while the procedure in use could obviously -
have been better, it was not reasonably within the licensee's control to have anticipated the tube snapping once it had been cleared of the reactor vessel. Contingencies had been in place for potential breakage in the vessel, but had been unexpected after that point.
Nonetheless, appropriate precautions and measures were taken for the remaining work.
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HUBBARD92-162 PDR
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We appreciate you informing us of your concerns and feel that we have been concerns. Should you have any further questions, or if I can be of funher assista matter, pleaw call me collect at (215) 337-5225.
S' cerely, 8
Edward Wenzinger, Chief Reactor Projects Bran
Enclosure:
Millstone 1 Inspection Repon 91-06.,
Section 4.2 bec:
Allegation fde, RI-91-A-0083 E. Conner E. Kelly W. Raymondfr. Shediosk,/
EG&G Idaho Representative (CARDONE\\ FROST)
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2 We appreciate you informing us of your concerns and feel that we have been responsiv Should you have any further questions, or if I can be of further assistance in this concerns.
matter, please call me collect at (215) 337-5225.
Sincerely, criain:1 Signed By Edward Wenzinger, Chief Reactor Projects Branch 4
Enclosure:
Millstone 1 Inspection Report 91-06, Section 4.2 bec:
Allegation file, RI-91-A-0083 E. Conner E. Kelly W. Raymond/T. Shediosky EG&G Idaho Representative (CARDONE\\ FROST) i i
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concurrences:
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NI RECORD OF ALLEGATION PANEL DECISIONS
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//li skcht PANEL ATTENDEES:
ALLEGATION NO.:
2.~1-9d-A-eo84 Chairman - 3. I LW s qiw s
=0 DATE:
644Y93 (Mtg.h2345)
Branch Chief -
Section Chief ( AOC) - E.M. /le,N, PRIDRITY:
High Medium Low SAFETY SIGNIFICANCE: Yes No Unknown Others T. S. Sieum,d-CONCURRENCE TO CLOSE0VT: DD BC SC (2 [. Mem e& $ o-f CONFIDENTIALITY GRANTED: Yes No (See Allegation Receipt Report)
IS THEIR A 00L FINDING:
Yes No y
IS CHILLING EFFECT LETTER WARRANTED:
Yes No l
HAS CHILLING EFFECT LETTER BEEN SENT:
Yes No HAS LICENSEE RESPONDED TO CHILLING EFFECT LETTER:
Yes No j
ACTION:
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5 Allegation File,(e) RI-91-A-0082,0084,0085,0096,0099,0Hsf T. Shediosky, SRI, Millstone J. Stewan E. Kelly M,R gm. A r
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION 5
REGION I I
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475 ALLENDALE ROAD f
KING OF PRUSSIA. PENNSYLVANIA 19404 JUL t 7 m j
Docket Nos. 50-336 f
File Number RI-91-A-0082 Northeast Nuclear Energy Company l
ATTN:
Mr. E. J. Mroczka l
Senior Vice President - Nuclear Engineering and Operations Group j
P.O. Box 270 I
Hartford, Connecticut 06141-0270 i
Dear Mr. Mroezka:
The U.S. Nuclear Regulatory Commission recently received information concerning activities at the Millstone Nuclear Power Facility, Unit 2. The details are enclosed for your review and
-l follow-up.
We request that the results of your review and disposition of these matters be' submitted to Region I within 30 days of the date of receipt of this letter. -We request that your response contain no personal privacy, proprietary, or safeguards information so it can be released to i
the public and placed in the NRC Public Document Room. If necessary, such information-shall be contained in a separate attachment which will be withheld from public disclosure.
The affidavit required by 10 CFR 2.790(b) must accompany your response if proprietary.
information is included. Please refer to file number RI-91-A-0082 when providing your ~
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response.
The enclosure to this letter should be entrolled and distribution limited to personnel with a "need to know" until your investigatica of the concern has been completed and reviewed by NRC Region I. The enclosure to this letter is considered Exempt from Public Disclosure in accordance with Title 10, Code of Federal Regulations, Part 2.790(a). ' However, a copy of this letter excluding the enclosure will be placed in the NRC Public Document room.
a The response requested by this letter and the accompanying enclosure are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511.
Your cooperation with us is appreciated. We will gladly discuss any questions you have -
l concerning this information.
3 Sine el
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les W. Iehl, D ector A)gWHy f
Division of Reactor Projects
Enclosure:
(10 CFR 2.790(a) Information)
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ENCLOSURE tIMITED-DIS,TRIBUTION 'NOTJOR.P_UJI:IC-DISCLOSHRE Issue 1:
The uiring diagrams insviving Reactor Coolant Pump RTD circuits have not been updatedfollouing modifications made under a PDCR to replace RTD circuit kmfe suitches with Weidmuller Test Blocks. Draning No. 25203-31069, Sheet 5, Rev. 3, dated August 29,1989, does not reflect the changefor at least 4 RTD circuits (TCD, TCC, TCA, TCB). The instrument loop diagrams (Draning No. 25203-28500, Sheets 140 & 146) show the Weidmuller Test Blocks. Also, in Draning number 25203-31069, Sheet 5, thejumpers shoun hetaren cable lead I and the cable shield ground on the loop diagrams are not shoun. In addition, access to the GRITS system, to venfy the latest drawing revisions, is restricted in ibat personal access codes are only validfor 30 days.
Request I:
Please discuss the validity of the above assertions. If discrepancies are found, please assess the significance of the discrepancies with respect to plant operation and safety and discuss any actions taken or planned to correct these discrepancies.
Issue 2:
The Steam Generator No. 2 mid-loop instnanentation (L-122) was not " operable" during drain-dounfor tube inspections on May 2,1991. GEM suitches werefound to be ' frozen" on in place. In addition, L-112 had an electronic noise problem caused by an improperly installedjumper. Thus licensee commitment that tuo monitors be operable during drain doun condition uns not being met.
Request 2:
Please discuss the validity of the above assertions. If any discrepant conditions are identified, please discuss their significance with respect to plant operation and safety during the Steam Generator No. 2 drain-down evolution. Also please discuss any actions taken or planned to correct these deficiencies.
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Issue 3:
Pressure indicating instrutnent (PI 6350 A/B and PI 6351 A/B) and mountingsfor service unter (SW) supply to cmergency diesel generators (EDG) are not seismically mounted. Any kind of shock uvuld be suficient to knock the gauge and valve of of the strainer. Additionally, the location of the taps as shown on the P&lD apparently does not coincide with the actual tap locations.
Request 3:
Please discuss the validity of the above assertions. If the assertions are valid, please discuss their effect on the safe operation of SW supply to the EDG. Please nrovide any actions taken or planned to ensure that seismic requirements for these instruments are being met.
Issue 4:
On May 3,1991, the Unit 2 Stack Radiation Monitor (luf 8132) was inoperable as a result of beingflooded with water. This monitor would have been inoperable anyway, as airflow had been isolated. Filling and pressure testing of Steam Generator (SG)
- 1 uns underway during the same time period. Problents with valve line-upsfor the rad monitor and the SG testing contributed to theflooding and monitor inoperability.
Additionally, health physics (HP) controls during removal of the waterfrom the monitor uns inadequate resulting in contamination ofpersonnel.
Request 4:
Please discuss the validity of the above assertions. If discrepancies are confirmed, please discuss actions that you have taken or will take to ensure that plant procedures regarding rad monitor operation, conduct of tests, and HP activities are being used properly.
Issue 5:
Procedure discrepancies exist between OP-2336E and SP-2617Afor the restoration of the line-up of the radiation monitor (RE-245), and its associated sample pump.
Operators routinelyfail to perform OP-2336E, Section 5.1, Step 5.1.13 which is to immediately close AOV-244A/B and AOV-245 when securingfrom condensate polishing facility discharges. Thisfailure tofollow procedures results in the sample pump to radiation monitor (RM-245) continuing to operate when the tank discharge is secured.
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v Request 5:
Please discuss the validity of the above assertions. If any discrepancies are identified concerning procedure noncompliance, please discuss their significance on the operation of radiation monitor RM-245. Please discuss any corrective actions taken or planned, to ensure operators are meeting procedural and technical specification requirements.
Issue 6:
Thefollowing discrepancies have been idennfied during an evaluation of Work Order AWO-M2-91-04411. These discrepancies idennfy continued non-compliance with procedures and poor response of operations and management to recurring problems with radiation monitor RE-245.
7he sample pump continues to run when the tank discharge stops at 15% tank a.
level (TK-11).
b.
The " Low Flow" switch does not always see a lowflow condition when TK-10 and TK-11 discharge pumps stop. The head of water in the pipe and tidal conditions afect theflow of water.
Operations normally rely on the 15% tank levelpump trip to stopflow causing c.
a lowflow to trip shut RE-245 discharge valve, and AOV-245. IfAOV-244A/B are shut and no lowflow condition exists, RE-245 sample pump will continue to ntn until AOV-245 is shut.
d.
Changes to OP-2336E were idennfied in 1989 to prevent the problems idennfled by AWO-M2-91-04411. However continued idennfied procedure non-compliance by operations has caused repeated problems.
Request 6:
Please provide an assessment of the above discrepant conditions. If the assertions are valid, please discuss their safety significance and effect on operation of radiation monitor RE-245. Please discuss any corrective actions that are being used to correct the problems.
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Docket Nos. 50-336 File Number RI-91-A-0082 Northeast Nuclear Energy Company ATTN:
Mr. E. J. Mroczka Senior Vice President - Nuclear Engineering and Operations Group P.O. Box 270 Hartford, Connecticut 06141-0270
Dear Mr. Mroczka:
The U.S. Nuclear Regulatory Commission recently received information concerning activities-at the Millstone Nuclear Power Facility, Unit 2. The details are enclosed for your review and follow-up.
We request that the results of your review and disposition of these matters be submitted to.
Region I within 30 days of the date of receipt of this letter. We request that your response contain no personal privacy, proprietary, or safeguards information so it can be released to the public and placed in the NRC Public Document Room. If necessary, such information shall be contained in a separate attachment which will be withheld from public disclosure.
The affidavit required by 10 CFR 2.790(b) must accompany your response if proprietary information is included. Please refer to file number RI-91-A-0082 when providing your response.
t The enclosure to this letter should be controlled and distribution limited to personnel with a "need to know" until your investigation of the concern has been completed and reviewed by NRC Region I. The enclosure to this letter is considered Exempt from Public Disclosure in accordance with Title 10, Code of Federal Regulations, Pan 2.790(a). However, a copy of this letter excluding the enclosure will be placed in the NRC Public Document room.
i The response requested by this letter and the accompanying enclosure are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork -
Reduction Act of 1980, PL 96-511.
Your cooperation with us is appreciated. We will gladly discuss any questions you have concerning this information.
- Sincqly,
-- s 1es W. kehl, Dhector C
Division of Reactor Projects.
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Enclosure:
(10 CFR 2.790(a) Information) j wm y
i O I - Stl 91 - %
ENCLOSURE Issue 1: The seismic qualification of pressure indicating instrument (PI 6350 A/B and PI
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6351 A/B) taps and mountings for service water (SW) supply to emergency diesel generators (EDG) has been questioned. Additionally, the location of the taps as shown on the P&ID apparently does not coincide with the actual tap locations.
Please discuss the validity of the above assertions. Please discuss any actions taken or planned to ensure that safe operation of the plant continues, Issue 2: Maintenance supervision has not adequately addressed the need for procedure improvements. Also, such improvements are not completed in a timely manner once initiated. Specifically, surveillance procedure SP-2410A, Rev. 5 was known to contain
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errors, but it was used for a period of approximately three months following identification of these errors.
Please discuss the validity of the above assertions. Please discuss any actions that you have taken or will take to ensure that necessary improvements to plant procedures are implemented in a timely manner.
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