Text

Responds to NRC Ltr Re Violations Noted in Insp Rept 50-002/98-202 on 980223-27.Corrective Actions:Temporary Operating Instruction to Lower Outlet Temperature Setpoint to 126 F Was Issued on 961008
	ML20217E390
Person / Time
Site:	University of Michigan
Issue date:	04/22/1998
From:	Fleming R; MICHIGAN, UNIV. OF, ANN ARBOR, MI
To:	Weiss S; NRC (Affiliation Not Assigned)
References
	50-002-98-202, 50-2-98-202, NUDOCS 9804270287
	Download: ML20217E390 (68)
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{{#Wiki_filter:r j ? g (, MICHIGAN MEMORIAL -PHOENIX PROJECT PHOENIX MEMORIAL LABORATORY FORD NUCLEAR REACTOR g ANN AMOR, MICHIGAN 48M2W m<, s e.,. l 50-002. To: Seymour H. Weiss, Director Non-Power Reactors and Decommissioning Project Directorate i Division of Reactor Program Management Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission PRESENTATION TO THE ENFORCEMENT CONFERENCE OF APRIL 22,1998 1.

SUMMARY

There are two apparent violations as a result of the inspection conducted at the Ford Nuclear Reactor on February 23-27, 1998. 1. Under 10 CFR 50.59 (a)(1) the licensee may make changes to g

V the facility without prior NRC approval only i f those changes do not l

involve changes in the technical specifications in our license. Apparent violation VIO 50-002/98202-01 points out that our April 1996 maintenance on the FNR primary pump resulted in a significant flow increase and therefore did require a TS change. This change in the TS had not been initiated at the time of the inspection. We do.not dispute the description of the events in the Inspection l Report under section 2.0 (1) "10 CFR 50.59 Review". When faced with a similar problem with the primary pump in April 1998, a proper 10 CFR 50.59 safety review was carried out with the benefit of the inspection report, the TS change identified, and License Amendment 44 was submitted to the NRC for approval. Until that approval is obtained, the modification request is approved by the SRC, and the modification is carried out, the FNR will not be operated in Forced Convection Mode. o , ) *5 \\k

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p 2. Our license (TS 6.6.2.b.2) requires us to prepare a written report to the NRC within 30 days should we discover a substantial variance from performance specifications contained in the technical j specifications and safety analysis. AlthouGh a problem was identified by FNR operations in October 1996, leading to a temporary reduction in outlet temperature set point on October 8,1996, and to the installation of a new LSSS rundown on inlet temperature in September 1997, no notification was made to the NRC. Apparent violation VIO 50002/98202-02 resulted from our failure to report the observed condition after it was identified in October 1996. l l O l l O 2

TABLE OF CONTENTS 1.

SUMMARY

ll. ROOT CAUSES AND MISSED OPPORTUNITIES FOR CORRECTIVE ACTION lli. CORRECTIVE ACTIONS TAKEN IV. SIGNIFICANT ISSUES AND LONG-TERM CORRECTIVE ACTIONS V. CONCLUSIONS Appendix: A CHRONOLOGY OF EVENTS IN REVERSE ORDER Attachments: A. FNR License Amendment 44, dated April 2,1998. This amendment received NRC approval on April 16, 1998. B. Minutes for FNR Safety Review Committee April 1,1998. (Draf t dated April 8,1998.) C. FNR memo " Observations During Primary Pump Replacement and Check Valve Removal" dated April 29,1996 D. FNR memo to "QA File for installing Standby Primary Pump" dated O Jenuary 27. 1987. E Administrative Procedure No. 201 " Ford Nuclear Reactor Modifications", proposed Revision 6, dated April 15,1998. (Rather than include a blank form, we have included FNR Modification Request No.129 for the current pump motor replacement, with its own attachments, as the example.) F. Annual Report on FNR Safety and Operations of the FNR Safety Review Committee (SRC) for the Office of the Vice President for Research of-the University of Michigan, dated August 26,1997. G External Auditor's Report, dated January 16,1997 with cover memo to the SRC dated April 21, 1997. H. FNR memo " Summary Report on the NRC Inspection of Fm Operations", dated February 27,1998. This memo, which was read by each member of the licensed staff, included a copy of Part 10 CFR 50.36. 1. NUREG-1138, Safety Evaluation Report for the FNR, July 1985. (Section 5) O 3

f i1. ROOT CAUSES AND MISSED OPPORTUNITIES FOR CORRECTIVE ACTION \\. - One root cause of the apparent violation stems from the fact that our TS provide a LSSS on core outlet temperature to protect a Safety Limit on core inlet temperature. This resulted in a situation where the protection was both power and flow dependent, and this was recognized at least as early as January 1987, see attachment D.The dependence on power is moot since the SL of 116 F only applies at 2 MW. The flow dependence means that the SL is less well protected as l the flow increases, even though core heat removal is enhanced. The fact that the analog temperature recorder allowed only a single set point meant that the temperature LSSS would have to be the highest temperature, that is the core outlet. The reason that the TS specify a Safety Limit on inlet temperature, rather than outlet temperature, is less obvious since the safety analysis seems to justify either limit. In any case, there was a missed opportunity for corrective action -in 1987 when the horizontal primary pump was installed. l Perhaps the first missed opportunity for corrective action had been in March 1985 when the new digital temperature recorder was Installed. It was now possible to have a LSSS on any individual O tempereture, end the modificetion tnet wee eventueiiv mede in September of 1997 to have an inlet temperature LSSS at 114 F could have been made with the new recorder installation in 1985, or at any time since. The key opportunity for corrective action came, of course, with the pump repair in April 1996. See Attachment C for a description of the work. The Part 50.59 review that we carried out in April,1998, which resulted our being required to change our TS, addressed the identical situation with the identical license that we missed in April 1996. The difference was the argument put forward in the Inspection Report in its critique of the 1996 10 CFR 50.59 Review. We missed the full significance of the problem until it was pointed out by the NRC.We believe that our technical judgment and actions were sound', but that our appreciation of the license requirements at the managerial level was deficient. Since I was personally involved in these decisions through regular discussions with the reactor manager during each step of the work, O perhaps I can indicate why this happened. In my own case, I fell into 4

I the mindset that, if the technical problem is low flow in the primary v \\ system, then more flow should be better. I did not adequately recognize that our TS have an implicit upper limit on primary flow. I Another example of this thinking was my not recognizing in September of 1997, when the rundown was installed at 114 F inlet l temperature, that a TS change was required. Here the mindset was l that, if we have satisfied all of the requirements of the TS, we l should be able to add additional features, such as a rundown, that will enhance the safe and proper operation of the reactor without violating our license. While this may be so in some cases, what I did not fully appreciate was that, if we take credit for the change as an LSSS, the modification must be reflected in the TS. ! am now clear l on both of these points. l l There were other opportunities for corrective action, s!nce the issue continued to come up. However, two opportunities were misee.1 when

1) Chris Berg, the Senior Operator, pointed out the problem to management on October 6,1996, and 2) when our external auditor, Ward Rigot, raised the question explicitly in his January 16,1997, l,_

audit report. In both cases we responded, with the Temporary id Operating Instruction lowering the outlet temperature setpoint to l 126 F on October 8, 1996, and with the Modification No. 123 l approved on June 30, 1997, to install the inlet temperature rundown. In neither case was the response sufficient. l While perhaps not a root cause, but a contributor to them, is the need to improve the timely flow of information to the FNR SRC. The l two changes which we have made to our Modification Request Procedure (AP-201) should help accomplish this. First, we make explicit in the request process when SRC approval is required prior l to the modification being carried out, and when SRC approval is required prior to unrestricted restart. Second, the requirement of two management approvals at key points in the 50.59 review should i greatly improve information provided to the SRC. In addition, as the l Director, I will in the future invite the Chair of the SRC, or his designate, to join me in the exit interview that I hold with our external auditor following the annual audit. 5

o 111. COPritECTIVE ACTIONS TAKEN U 1. A Temporary Operating Instruction (TOI) to lower the outlet temperature setpoint to 126 F was issued on October 8,1996. I 2. A core inlet temperature rundown at 114 F for power greater I than 1.6 MW as per Modification Request No.123 was installed in September 1997. 3. The FNR ceased operation in Forced Convection Mode on March 26,1998, pending approval of Amendment 44 to the License and the decision to return to forced convection operation. 4. The FNR SRC met on April 1,1998, to evaluate the 10 CRF 50.59 review of Modification Request No.129, which resulted in License Amendment 44 being submitted to the NRC to add a LSSS on the inlet temperature of 114 F at 2 MW, on April f, 1998. This amendment formalizes the rundown installed in Septe:nber 1997 as per Modification Request No.123. We received approval for this amendment from the NRC on April 16, 1998. The Amendment is C) included as Attachment A. 5. A new T1 temperature sensor has been installed at the entrance to the holdup tank, just after the downcomer. Temperature sensor T2 is located downstream in the primary, after the holdup tank and the primary pump, but ahead of the heat exchanger. During steady state operation we observe no difference in reading between these two temperatures; both accurately represent the FNR core exit temperature. 6. We have prepared Revision 6 to Procedure AP-201, " Ford Nuclear Reactor Modification", dated April 15, 1998, which a) provides two levels of SRC prior approval of modifications, prior to implementation, and prior to unrestricted restart; b) makes explicit l the Health Physicist involvement in the ALARA review process; and l c) will require approval and signoff by both the reactor manager and an assistant manager. The revised procedure requires approval of the SRC, and since it is part of our QA Plan, approval of the NRC. In the interim we will use both Revision 5 and Revision 6 in parallel. We O have used this AP-201 (Rev 6) to generate a new Modification i 6 i i

p Request No.129, including its 10 CFR 50.59 review. This is included i v as attachment E. The requirement of two managers' approval at key points in AP-201 will result not in just a second column of initials, j but rather will provide a dialogue, which should prevent the sort of oversights we have made in the past. 7. As pointed out on page 6 of the Inspection

Report, we committed to prepare a copy of 10 CFR 50.36 for required reading by all licensed operators. This was done during the week of the inspection as a memo to licensea staff entitled " Summary Report on the NRC Inspection of FNR Operations", dated February 27,1998, in addition, the Director prepared a memo for the SRC members entitled " Tech Spec Definitions of ' Safety Limits' and ' Limiting Safety System Settings'" and dated March 23,1998.

Copies of these memos, which were also provided to

staff, are included as attachment H.

8. We will incorporate explicit training of licensed staff on the definitions and cignificance of Safety Limits. This will include current operators, operator re-training, and training of new q V operators. For the record, I will state here the key result of the FNR Safety Analysis, dated November 1984, which will be part of that training. Using a model of the FNR which is clearly conservative both physically and calculationally, we know that even with the four forced convection reactor parameters at their safety limit values - 4.68 MW, 900 gpm,116F, and 18 feet depth - the maximum f uel l cladding temperature will be less than 235 F, the saturation temperature of water at that depth, everywhere in the core. In order for the cladding temperature to exceed 235 F, at least one of these l parameters must violate its safety limit. A physical description of the FNR is provided as Attachment 1. l l O 7

IV. SIGNIFICANCE OF THE ISSUES AND LONG-TERM CORRECTIVE ACTION I believe that the safe operation of the FNR is not in question. However, the issues raised in this hearing are crucial to the continued operation of the Ford Nuclear Reactor in a way that is clearly consistent with its license. The mismatch between our Safety Limits and our LSSS requirements to protect them has caused i us uncertainty and confusion since the beginning. It has led to an awkwardness in carrying out our work as we wrestle with the illogic, and now it has contributed to the current apparent violation. After a careful study of the safety analysis which underlies them, I am convinced that it will support a Safety Limit on core outlet temperature, rather than core inlet temperature. Although amendment 44 addresses the license requirement that the LSSS settings in our technical specifications must always protect our Safety Limits, I believe that to understand the root cause of the problem we need to re-analyze the Safety Limits for Forced Convection Flow. We have used the heat transfer model of our Safety Analysis to generate one-dimensional calculations to confirm the O noint-reector engroximetion used to determine the eefety iimite in our TS. As expected, the results in the TS are always conservative and quite accurate. Using these calculations, we have initiated a study of a new set of Safety limits and LSSS settings for Forced Convection Flow operation of the FNR with the goal of Safety Limits on 1) reactor power, 2) primary flow rate, 3) pool height, and 4) the core outlet temperature. This will have major advantages, in addition to removing the logical flaw that has caused such problems. First, the core outlet temperature is sensitive to reactor conditions well before the inlet temperature is, and therefore provides a better safety limit for heat removal. Second, we have installed a temperature sensor in the primary system at the entrance of the holdup tank to give a core outlet temperature measurement af ter mixing but before any significant heat loss. Preliminary results from this study have not turned up any technical or license reasons to doubt that this approach is possible. In the meantime, our License with Amendment No. 44 will maintain safe and proper operation of l the FNR. l O l 8 l 1

V. CONCLUSIONS in our March 24, 1998, inspection Report No. 50-002/98202 the Director of the Division of Reactor Program Management, Jack W. I Roe, expressed three concerns: 1) a lack of understanding of our license requirements,

2) an incomplete implementation of our modification program, and 3) an ineffective implementation of our reporting requirements.

I believe that by our actions to date we have shown that we now do have an understanding of the license requirements, and by our commitrnent und the corrective actions we have in place the sin ture to carry out the full implementation of our responsibilities under the license. ]& i r Ronald F. Fleming, Director O uichioen uemoriei-ehoenix eroject j I l 1 I \\ O 9 j

1 APPENDIX 7(d A CHRONOLOGY OF EVENTS IN REVERSE ORDER 1 Aoril 16. 1998 NRC approval of Amendment 44 was received. (Procedures AP-201, CP-206, OP-101, OP-103 and OP-106 have been revised to reflect amendment 44.) Aoril 2.1998 - Amendment 44 to the Ford Nuclear Reactor License submitted the US NRC. (A copy is included as Attachment A.) Aoril 1. 1998 - The FNR Safety Review Committee met to discuss inspection Report and Amendment 44. (A copy of the proposed minutes is included as Attachment B. Note - Until the SRC approves its minutes at the next meeting, this version remains a draft.) March 25.1998 - The FNR experienced the failure of the primary coolant pump motor. (A Part 50.59 review for the installation of a spare pump resulted in an apparent Unresolved Safety Question as defined in Part O 50.59(e><2)<iii) beino reised. it wee conciuded tnet Amendment 44 wee required before the 50.59 review could be completed and the primary pump repair executed. The FNR has ceased operation in forced convection mode pending completion of this work. (As an aside, upon tearing down the pump

motor, FNR staff concluded that the reason for its failure was a manufacturing defect which did not allow full lubrication. When the motor was returned to the vendor, they agreed with our findings and are providing us with a replacement motor at no charge.)

February 23 27. 1998 - Routine Safety inspection conducted by m C Inspector Tom Burdick. (This inspection resulted in inspection Report No. 50-002/98202(DRPM), dated March 24, 1998.) Seotember 30. 1997 - Installed inlet temperature rundown on T5 of 114 F as per Modification Request No.123. (This instaliation will give a reactor rundown whenever the FNR is at greater than 80% of full power and the T5 temperature exceeds 114 F.) Aoril 29. 1997 - SRC discusses LSSS set point issue, and on June 30, h-1997, approved Modification No.123. I

1 November 11-13. 1996 - Audit of facility carried out by Ward Rigot. (The (] auditor's report, dated January 16, 1997, and FNR management response to it, dated April 21, 1997, are included as attachment G.) October 8.1996 - Temporary Operating Instruction issued to lower the set point for outlet temperature to 126 F. October 6.1996 - Senior Reactor Operator questions whether the Safety Limit of 116 F on inlet temperature is adequately protected by the LSSS of 129 F on the outlet temperature, given the current primary flow rates. Aoril 18-29. 1996 - Primary coolant pump and motor replaced and check valve ; removed as per Modification Request No.120. ( Due to low primary flow a system repair was initiated which resulted in the replacement of a primary pump with a 20 horsepower motor with one using a 25 horsepower motor. A memo " Observations During Primary Pump Replacement and Check Valve Removal" dated April 29,1996 is included as Attachment C.) Fall 1983 - New Fybroc pump with a 25 horsepower motor installed in the secondary system. Auoust 1987 - Installation of new Fybroc horizontal primary pump was carried out, increasing flow from about 1000 gpm to 1130 gpm. ( A memo to the QA File for Installing the Standby Primary Pump dated January 27, 1987, included as Attachment D, discussed the safety aspects of primary flow rate.) l Julv 29.1985 - Current FNR Operating License and Tech Specs issued. (With this license the power versus primary flow operating curves were replaced by the four Forced Convection Safety Limits of 900 gpm flow, 4.68 Mw power,116 F Tin and 18 feet pool height.) i March 18.1985 - Instrument upgrade allowed replacement of old analog temperature recorder with a new digital recorder. ( All of the temperature data T1-T12 is recorded on a single stripchart recorder, from which the signal is taken for the rundown function. The analog recorder had a single rundown setpoint, which was triggered by any one of the temperatures, that is, the highest one. The digital recorder has set point capability f or (3 each temperature individually.) %) r 2 i

The University of Michig:n 4 Michigin Mem:rirl Phoenix Prsject Office of the Director 2301 Bonisteel Boulevard Ann Arbor, Michigan 48109-2100 April 2,1998 Docket 50-2 License R-28 United States Nuclear Regulatory Commission Document Ccatrol Desk 1 Washington, D.C. 20555 i Re: Ford Nuclear Reactor License and Technical Specifications, Amendment 44: Forced Convection Inlet Temperature LSSS and Reactor Coolant Temperature LSSS The University of Michigan hereby submits Amendment 44 to the Ford Nuclear Reactor Technical Specifications. The changes are explained below. Enclosed is a copy of the revised pages with the changes underlined and denoted in the right margin. On March 25,1998 the FNR experienced a failure of the primary coolant pump. The safety system (High Power /No Flow scram) operated as intended and the reactor was safely shut down. A spare pump is available on site. After performing a Part 50.59 review for installation of the spare pump an apparent Unresolved Safety Question as defined in Part 50.59(a)(2)(iii) may exist. We have ceased operation of the reactor in forced convection cooling mode until this issue is resolved. The issue involves using Rextor Coolant Outlet Temperature as the LSSS caannel to protect against exceeding the Safety Limit on Reactor Coolant Inlet Temperature. The current Reactor Coolant Outlet Temperature LSSS of 129oF (which includes a 20F margin for measurement error) is based on the 150F AT that would exist across the core at the Safety Limit coolant inlet temperature of 116oF and Safety Limit coolant flow rate of 900 gpm while operating at 2 Mw thermal power. The Safety Analysis Report shows that increasing flow above 900 gpm increases core cooling. Improving the cooling gives a larger margin of protection against the fuel cladding reaching the boiling point of the water coolant, provided the Safety Limit on the inlet temperature of 116oF is satisfied. And, boiling is the event these Safety Limits are intended to protect against. I However, forced convection operation at any flow rate above the Safety Limit value of 900 gpm causes a decrease in the margin between the coolant temperature LSSS and Safety l Limit because increasing the flow rate decreases the AT at 2 Mw. The LSSS on coolant exit temperature becomes ineffective at preventing the 2 Mw coolant inlet temperature Safety Limit of 1160F from being exceeded. This amendment incorporates into the Technical Specifications a safety channel on reactor l coolant inlet temperature that was installed to address this problem on September 30,1997. l The LSSS causes an auto rundown at an indicated core inlet temperature of 1140F when the 1 ON l

l 2 l l l pd reactor is operated in forced convection mode at thermal power levels greater than 1.6 Mw as indicated on the Log N - Period Safety Channel. This safety channel provides automatic protective action to prevent exceeding the 2 Mw reactor coolant inlet temperature Safety Limit of 1160F under all anticipated normal and adverse reactor operating conditions. Additional minor corrections have also been made to the Technical Specifications. The following is a brief description of the specific changes: Page 2 of License R Condition C. (2). This change identifies Amendment 44 as the current Technical Specification l revision. Page 7 of the Technical Specifications. Condition 2.2.1 Applicability, Reactor Coolant Inlet Temperature Identified Page 8 of the Technical Specifications. l New condition 2.2.1.2 adds the inlet temperature LSSS at i140F for thermal power l levels greater than 1.6 Mw. Page 13 of the Technical Specifications, Table 3.1 REQUIRED SAFETY CHANNELS i l Function of the Log N Period Channel is modified to indicate that safety related power level interlocks are derived from this channel. j Reactor Coolant Inlet Temperature Channel is added which initiates an auto rundown at i143F for power levels above 1.6 Mw. I Reactor Pool Temperature Channel is added to specifically identify the Safety Channel that is required for the reactor pool temperature during Natural Convection Flow Mode. This Safety Channel initiates an Auto Rundown at pool temperatures greater than 1290F when the reactor is operated in natural circulation cooling mode. I Page 17 of the Technical Specifications, Section 3.2 Reactor Safety System Bases. Changed to identify the purpose of the Coolant Inlet Temperature and Reactor Pool Temperature Auto Rundowns. Changed to correctly identify the Pool Level Auto Rundown Channel as providing automatic protection against low pool level rather than operator action. Sincerely, l' Ronald F. Fleming. Director Michigan Memorial-Phoenix Project I cc: Theodore Michaels. USNRC Project Manager Thomas Burdick USNRC Region III j i 1 l l j

i >pERATING LICENSE AND TECHNICAL SPECIFICATIONS Fora Nuclear Reactor Docket 50-2. License R-28 Amendment 44: proposed C. This license shall be deemed to contain and is subject to () the conditions specified in parts 20, 30, 50, 51, o5, 70, and 73 of 10 CFR Chapter I, to all applicable provisions of the'Act, and to the rules, regulations and orders of the Commission now or hereafter in effect and to the additional conditions specified below: (1) Steady State and Maximum power Levels l39 l The licensee is authorized to operate the facility at a steady state power level of 2.0 megawatts (thermal). The maximum power level shall not be in excess of 2.2 megawatts (thermall. (2) Technical Specifications The Technical Specifications contained in Appendix A. as revised through Amendment No. 44, are hereby [44 incorporated in the license. The Ticensee shall operate the facility in accordance with the technical specifications. (3) physical Security plan The licensee shall fully implement and maintain in j37 l (/~T,,) by the Commission and all amendments and changes made effect all provisions of the physical security plan approved l pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p), respectively. The approved plan, which is exempt l from public disclosure pursuant to the provisions of 10 CFR

2. 790( d )( 1 ), is entitled " Ford Nuclear Reactor Security i

l plan," Revision 1, submitted by letter dated May 17, 1991, i and supplemented February 17, 1992. D. This license is effective as of the date of issuance and shall expire twenty years from its date of issuance. l FOR THE NUCLEAR REGULATORY COMMISSION l Director Division of Licensing Date of Issuance: July 29, 1985

Enclosure:

Appendix A, Technical Specifications i O-

, c ar..u _.m c t c.s e t A.ND TECHNICAL SPECIFICATIONS Ford Nuclear Reactor Docket 50-2. License R-28 Amendment 44: proposed 2.1.2 Safety Limits in the Natural Convection Mode I) Applicability: This specification applies to the interrelatea variables associated with core thermal and hydraulic performance in the natural convection mode of operation. These variables are: Reactor Thermal power, p Reactor Coolant Inlet Temperature, Ti Height of Water Above the Top of the Core, H Objective: To assure that the integrity of the fuel clad is maintained. Specification: 1. The true value of the reactor thermal power (p) shall not exceed 380 kw. 2. The true value of the reactor coolant inlet temperature (TI) shall not exceed 131 oF. 3. () The height of pool water above the core (H) shall not be less than 18 feet. Bases: The basis for_ natural convection safety limits is that the calculated maximum cladding temperature in the hot channel of the most compact FNR core (25 elements) will not reach the boiling point of the water coolant at a depth of 18 feet. 2.2 Limiting Safety System Settings (LSSS) 2.2.1 Limiting Safety System Setting in the Forced Convection Mode Applicability: This specification applies to the set points for the safety channels monitoring reactor thermal power ( p), primary coolant flow (m), height of water above the top of the core (H), l44 core exit temperature (T.) and core inlet temperature (T1). O page 7 ilI e

vren.441.u Litt.\\de A.Nu IECHNICAL SPECIFICATIONS Ford Nuclear Reactor Docket 50-2 License R-28 Amendment 44: Proposed Objective: (~T \\- To assure that automatic protective action is initiated to prevent a safety limit from being exceeded. Specification: } l 1. The limiting safety system settings for reactor thermal power (P), primary coolant ( flow through the core (m), height of water j above the top of the core (H), and reactor j coolant exit temperature ( Te ) shall be as follows: Variable LSSS 1 P (Max) 2.60 Mw l39 m (Min) 900 gpm H (Min) 19 ft T (Max) 129 F

2. The limiting safety system setting for l44 reactor coolant inlet temperature at tnermal

('i power levels above 1.6 Mw shall be: (_j Ti (max) = 114 oF l44 Bases: The limiting safety system settings for forced convection assure that automatic protective action will correct the most severe abnormal situation anticipated before a safety limit is exceeded. l 2.2.2 Limiting Safety System Settings in the Natural Convection Flow Mode l Applicability: These specifications apply to the setpoint for the safety channels monitoring reactor thermal power ( p ), pool water level (H), and pool water l temperature (T). Objective: To assure that automatic protective action-is initiated to prevent a safety limit from being exceeded. i Page 8

1 OPERATING LICENSE AND TECHNICAL SPECIFICATIONS Fora Nuclear Reactor Docket 50-2 License R-28 Amendment 44: Proposed ) Table 3.1 REQUIRED SAFETY CHANNELS Minimum Number Channel Setpoint Reauired Function Log Count Rate 2 cps 1 Rod Withdrawal Interlock Log N Period 1 Wide range power level, power level l44 interlocks, and input [ for period scram Period Safety 5 see 1 Scram Level Safety-122.5% (2.45 Mw) 2 Scram l39 High Power /No (a) 900 gpm 1 Scram 1100 kw Water Flow (b) holdup tank isolation valve not fully open (c) holdup tank Ox static pressure 1 psig below full power value High Power / Header Down 1 Scram 1100 kw Header Up/ 900 gpm 1 Scram No Water Flow l Building Exhaust 1 mrem /hr 1 Scram Radiation Level Buildihg Alarm 1 Scram Manual Switch Manual Scram Switch 1 Scram Magnet Power Keyswitch 1 Scram Reactor Coolant l44 Inlet Temp. 114 oF 1 Auto Rundown 1 1.6 MW l Reactor Coolant Exit Temp. 129 oF 1 Auto Rundown (j~'\\ \\ Reactor pool 44 Temperature 129 oF 1 Auto Rundown Page 13

OPERATING LICENSE AND TECHNICAL SPECIFICATIONS Ford Nuclear Reactor Docket 50-2, License R-28 Amendment 44: Proposed Il-Power-flow coincident scrams provide redundant channels I to assure that an automatic loss of flow scram will f occur in the event of a loss of flow when the reactor is operating at power levels above 100 kw. j i The rod withdrawal interlock on the Log Count Rate l channel assures that the operator has a measuring channel operating and indicating neutron flux levels during the approach to criticality. l The reactor coolant inlet temperature, reactor coolant l44 ) exit temperature, reactor pool temperature auto l rundown functions and the pool level auto rundown assure that the reactor will not be operated above the safety limit for core inlet temperature and below the safety limit for pool level in either natural or 44 forced convection mode. The manual scram button and the magnet power keyswitch provide two methods for the reactor operator to manually shutdown the reactor if an unsafe or abnormal l condition should occur and the automatic reactor protection does not function. The use of the area radiation monitor system assures r that areas of the facility in which a high radiation area could exist are monitored. Specifications 3.2.2 and 3.2.3 assure that the safety system response will be appropriate. Mechanical holddown devices specified in 3.2.4 ensure 38 that a fuel element cannot be lifted from the core by withdrawal of a control rod and be released back into the core resulting in a sudden positive reactivity insertion. 3.3 FNR Confinement Building Applicability: This specification applies to the Ford Nuclear Reactor confinement building requirements. Objective: l To assure that the Ford Nuclear Reactor Building 38 confinement integrity is maintained during reactor operation and to minimize the release of airborne radioactive materials from the reactor building. Page 17

SAFETY REVIEW COMMITTEE MEETING MINUTES TO: Safety Review Committee Members: Professor Dale Briggs Professor Massoud Kaviany Professor Henry Griffin Professor John Kine Professor John Lee! Chairman Professor James Martin Professor Fredrick Neidhardt Mr. Mark Driscoll Mr. Douglas Wood i Professor Ronald Fleming, Director Philip Simpson, Assistant Manager FROM: Bemard P Ducamp i Assistant Manager, Reactor Operations i DATE: April 8,1998

SUBJECT:

Agenda: First SRC Spring 1998 Meeting FNR Operation in Forced Convection. Safety Evaluation for a Proposed License Change Building: Phoenix Lab Room: FNR 3102 r Date: April 1,1998 L Time: 15:00 Attendees: Safety Review Committee Members: Professor Henry Griffin Professor Massoud Kaviany Professor John Kine Professor John Lee! Chairman Professor James Martin Professor Fredrick Neidhardt Mr. Mark Driscoll. RSO Mr. Douglas Wood Professor Ronald Fleming, Director Philip Simpson, Assistant Manager Bemard Ducamp, Assistant Manager Heath Downey, Health Physicist Chris Brannon, Sr. Electronics Engineer Michael Hartman, Eng. Tech. III The topic of discussion for this meeting was proposed license amendment #44, a proposed change that l resolves an unreviewed safety question regarding an unprotected Tin =116 F Safety Limit in FNR license R-28. O ^aaitie iiv. e<*er eis irice t ie ees -iii #eee revie 8 the sac this ePries. i ositiee te the #ermei 7 Tagelof 5

1 topics. It was decided to schedule multiple SRC meetings. The intention was to keep individual sessions more focused. This was deemed a much better alternative than one gigantic meeting lasted into l ungodly hours of the night. t 4 V John Lee, Chairman, opened the meeting by having everyone introduce themselves. Then Professor Massoud Kaviany, a new member, was introduced. Ron Fleming gave the first group of presentations. Ron Fleming referred to the March 24,1998 NRC inspection report (Ref. 3) for FNR. This report identified two apparent violations:

1. the failure to conduct a comprehensive safety review as required by 10 CFR 50.59; and

2. the failure to notify the NRC as required by FNR Tech Sp:cs when an inadequate LSSS was j

identified in accordance with the Safety Analysis Report analysis. Modification Request #120 (Ref. 5) - for a primary coolant pump changeout in April 1996 - failed o identify an unresolved safety question (USQ), specifically that the Tin =ll6F (max) @ 2 MW SL was not protected by the associated Texit=129 F(max) LSSS. Ron Fleming presented the SRC with a review of FNR's SL's and LSSS's for fo. reed convection flow (Ref. 4). Ron Fleming then gave an introductory presentation on an analysis entitled "FNR Core Thermal Hydraulics" (Ref.1) which has been completed. This analysis is a straightforward but thorough development of the temperature profiles for a 25 element FNR core. This analysis accounts for core heat generation in all three directions (x,y,z) and represents heat removal by the primary coolant in the direction of flow (z) in the hot channel. Some of the results of the analysis were presented, showing that the current SAR is still adequate and conservative. (d John Lee then asked Bernard Ducamp, Assistant Manager for Reactor Operations, to review with the 3 SRC reference materials #2 to 8. Reference 2 was described as a system description used for operator training and licensing. Bernard Ducamp said that, as a rough approximation, SRC members could think of the relationship between primary coolant flow and core delta-T as follows: Flow (gpm) Core Delta-T (F) 900 15 1000 14 1100 13 1200 12 This demonstrates the inverse relationship existing between flow and core delta-T when operating the reactor at a constant 2 MW of thermal power. In reference 3, pages 3 to 5 described the NRC inspector's findings, and page 6 the two apparent violations. Doug Wood emphasized the importance of appreciating NRC concerns with programmatic deficiencies in the modification request esd s afety analysis processes. Ron Fleming absolutely agreed with Doug Wood. The current modification e cpests and proposed license changes being performed (Refs. 8,9) will point out areas of deficiency in the older and non-comprehensive modification requests and safety analyses (Refs. 5.6,and 7) In reference 4, it was pointed out that the SAR and the core thermal hydraulics analysis (Ref.1) show q that boiling in the core is approached when all 4 process variables (power, flow, pool height, and inlet V temperature) are at the SL values. ( Tage2cf5

Reference 5 is not immediately relevant to core thermal hydraulics, but does review the associated check valve issues. Reference 5 does indicate that the original primary coolant system design was for 3000 p gpm total. O \\ Reference 6, the primary pump changeout in 1996, was reviewed. Doug Wood pointed out that the I associated safety evaluation for this modification was deficient. Doug Wood also pointed out the J importance of properly understanding what constituted " substantial changes " requiring prior SRC approval. It was noted that on the first page of this modification request, block 4 - Tech Spec l Change..Yes/No and block 5 - NRC Approval..Yes/No were both marked No. John Lee pointed out that in the early-mid 1980's (when the SAR and current reactor license were written), the temperature recorder was an older unit capable of only providing one rundown signal from the highest channel. There was no choice but to use Tout, the highest temperature in the primary coolant system during forced convection flow, as the LSSS to protect the Tin SL. i Reference 7, the modification request which installed the Tin =114 F @ 2 MW rundown in Sept 1997 was then reviewed. It was noted that the safety evaluation section on page 1 of the modification request indicated the unfortunate decisions that a Tech Spec Change and NRC approval were not required. Fred Neidhardt suggested that a Yes/No block 3.1 SRC Approval - Before Modification, should be i followed by a Yes/No block 3.2 SRC Approval - Before Reactor Startup. This suggestion was approved j by SRC members. Procedure AP-201, " Ford Nuclear Reactor Modifications" will be revised to include this. For the SRC review of this modification,it was noted that on page 4 of 8 the SRC Meeting Minutes for April 29,1997 stated the following. O "The SRC also considered the alternative of reducing primary coolant flow rate by throttling the pump discharge valve instead of performing this modification. This modification was judged preferable to reducing primary coolant flow rate." This indicates the SRC understanding at that time to be consistent with the Safety Analysis Report - more flow is better (provides more core cooling). The USQ lurking m Tech Specs was not noted. Reference 8. the modification request for installing a 20 HP pump to replace the failed 25 HP unit (motor bumout) was then reviewed. Mark Driscoll, RSO, inquired about the ALARA review. He was concerned about reports of personnel contamination during the work in the basement. Two FNR licensed reactors picked up 1000 counts per minute (frisker) contamination on their clothing while performing very physical work. They were wearing protective clothing, and some of the contamination appears to have occurred due to pool water soaking though the outer layer of protective clothing. It should be noted that the reactor basement is a very warm and humid work environment, with limited overhead clearance. The contaminated clothing was either given to HP (hold until decayed) or the contaminated parts (knees of trousers) were cut off and discarded as radwaste. One Plant Department electrician, Joel Foos, picked up 1000 counts per minute on the toes of his shoes. The electrician passed through a hand and foot counter along with Bernard Ducamp upon leaving I the basement without setting off an alann. Bernard noted some moisture on Joel's boot tops. Joel was wearing booties during his work in the basement. Apparently, the toes of the booties wore through at some point. The boot tops were checked using a frisker, and Joel then went to see the HP Tech to address the contamination on his boot tops. Joel left his shoes at Phoenix with the understanding that he h could get them back after the contamination decayed away. v T.yeJ of 5

FNR's HP staff is preparing a full report. (V ) Mark Driscoll, RSO, specifically requested that future ALARA reviews include a Health Physicist, not a H.P. technician. Mark Driscoll requested that a separate sign-off block be added after the ALARA l review block to remind FNR managers to involve a Health Physicist in the planning stages of facility modifications. The SRC requested that FNR personnel re-perform page 2 of the modification request checklist, paying careful attention to each item. The SRC was not concerned with the adequacy of the work, but rather with the thoroughness of the documentation. The SRC asked to see this page 2 again after FNR staff's review. The SRC also requested that the associated safety evaluation be rewritten to indicate that a change to Tech Specs is required: the proposed license change (Ref. 9) The SRC was then ready to review the proposed Amendment 44 to FNR License and Tech Specs (Ref. 9). SRC made a few suggestions for the wording of a few phrases. The SRC asked that the letter clearly indicate that the addition of the Tin =114 F @ 2 MW mndown was physically installed in 1997. This license change is written to legitimize it as a LSSS, as it should have been done in 1997. The SRC then voted to unanimously approve proposed Amendment 44. Amendment 44 was finalized by Phil Simpson and FNR staff, then submitted to the NRC on April 10, 1998 (Ref.10). John Lee asked the other SRC members if there was any member had any objections or concerns in i operating the reactor at 50 KW (max) in natural convection, knowing that the LSSS is at 100 KW. All SRC members indicated they were quite satisfied that 50 KW (max) was acceptable. (It should be noted that the low power operations performed after the primary pump change out and the discovery of O the USQ in March 1998 were at 20 KW (max).) V Reference Materials I to 9 were already distributed to all members of the SRC prior to the April 1 meeting. They are therefore not attached to these minutes in an effort to be environmentally friendly. Reference 10, the final submittal of Amendment 44 to the NRC, was prepared after the meeting. A copy of Reference 10 is enclosed with this distribution of the meeting minutes. REFERENCE MATERIALS

1. FNR Core ThermalHydraulics

2. Primary Coolant System Description.

3. NRC Inspection Report #50-002/98202 (DRPM)

4. Letter from R. Fleming to J. Lee, dated March 23,1998

5. Letter from R. Fleming to T. Michaels, NRC, " Primary System Check Valves " dated June 18,1997.

6. Modification Request #120, " Primary Coolant System: #2 Pump / Motor Changeout and Check Valve Removal."

7. Modification Request #123, " Install Rundown on T5 (Temperature System) at 116 F @ 2 MW."

8. Modification Request #129, " Install a Pump Driven by a 20 HP Motor into the Primary Coolant System (Replacement for the failed 25 HP pump)."

9. FNR License and Tech Specs: Amendment #44, as proposed.

p

10. FNR License and Tech Specs: Amendment #44, as submitted to NRC.

G Tage 4 of 5

J xc: Ron Fleming Heath Downey Phil Simpson O Rob Blackburn Reactor Operations Personnel i i O O Tage 5of 5

l O L) April 29, 1996 OBSERVATIONS DURING PRIMARY PUMP REPLACEMENT AND CHECK VALVE REMOVAL Periodic decreases in primary flow from near 1000 gpm to as low as 700 gpm were first thought to be caused by a problem with primary coolant pump 2 impeller or the flow sensing lines. The installed 20 HP pump was replaced with a pump that was identical mechanically, but that has a 25 HP motor. The sensing lines were flushed and blown out. The reduction in flow problems reoccurred. An inspection of the primary piping internals was started from the holdup tank forward. A check valve on the discharge of the pump was found to be broken. The valve seat was loose. It would periodically flop into the outlet opening causing a major reduction in primary flow. The internals of the check valve were removed. The purpose of the check valve was to prevent back flow if the two primary coolant pumps are run in parallel. This is .ever done. Since removal of the check valve and installation of the new pump, primary coolant flow has been steady near 1100 gpm. (ml Pump Replacement \\s) "~ Pump replacement was relatively easy. It was accomplished by two operators in about two hours. Horizontal pumps offer a strong advantage over vertical pumps for ease of replacement. Check Valve The check valve in the pump loop was installed to prevent back flow if pumps 1 and 2 are run in parallel. The two are never run in parallel. The check valve simply served as an unnecessary impedance to primary flow. Butterfly Valve The butterfly valve from the reactor pool into the holdup tank held very well. The valve was shut and the portion of the primary system with the holdup tank was drained. It was allowed to sit over night before the piping was opened for examination. There was no indication of leakage or decrease in pool level. When the Inspection cover was opened, one bolt was left in place in case leakage did develop so the cover could be rapidly swung back in place and sealed. Holdup Tank ! O) (- Part of the inside of the holdup tank was inspected. It had been preserved over twenty years ago with iron oxide (red) paint. The paint appeared almost new. No sign of any deterioration was observed.

I i i Primary Piping The rubber lining in the primary suction piping was inspected. It showed no sign of deterioration. The lining appeared to be tenaciously attached to the steel pipe. The lining is quite rough j which does add to the friction loss in the pipe, but that appears to be the way it.was applied. Primary Coolant Flow Sensing Lines Primary coolant flow has two independent sensors, but they both tap off the same set of sensing lines. A second set of sensing taps is available. Unless something unusual is discovered in the process, the sensors will be rearranged so that each is attached to a different tap. It was observed that the taps were reasonably dirty inside. They were flushed and blown out. This will be set up as an annual maintenance item. It may be wishful thinking, but the flow rate seemed to fluctuate less once the lines were. flushed and blown, i 1 1 (-) 1 O) l l

J UNIVERSITY OF MICHIGAN Phoenix Memorial Laboratory i Ann Arbor, Michigan Memo to: QA File for Installing Standby Primary Pump j From: Gary M. Cook l Quality Engineer Date: January 27, 1987

Subject:

Safety Analysis of Modifications The installation of the standby primary pump is to be a parallel l installation of an exact replacement for the current primary pump. l The purpose is to allow initially for a repair of the original, 4 l and still continue to operate a normal reactor schedule. Once l both pumps are available for use, the operations crew will have the option of using either. The new equipment will be selected to be the same as that which is being duplicated. This includes the pump, two isolation valves, a () check valve, and a temperature sensor. These are all standard off-the-shelf parts ordered from the manufacturer. The flow rate in the primary system is intended to be the same as original, approximately 1050 gal / min. During the last few years the flow has dropped from this value to about 960 gal / min because of pump wear. Once the new pump is operational the original will be brought up to like-new specs so that it will also pump 1050 gal / min. As long as the primary flow remains at 1050 gal / min maximum, there is reason to believe that all previous safety analyses are still valid and that there are no new situations to be analyzed. There will be administrative controls, including written procedures, designed to prevent the simultaneous operation of both pumps. Electrical switchgear is being designed which will not permit the second pump to be started if one pump is running. Given the above, it is concluded that reactor operation will not be significantly affected with the installation of the standby primary pump and that there will be no new safety related issues. This installation has not changed the probability of an occurrence, nor will it increase the consequences of an accident. It will not create a different type of malfunction or accident and O the margin of safety in tech specs has not been decreased.

~ q AP-:01 Fera Nuciear Reactor Modifications Revisioni6LO41598 FORDNUCLEAR REACTOR MODIFICATION REQUEST NO. l / 2 9 l Page 1 of 3 l > w..a .1 u nc,a o e 2 u ~ ,.,e o er, _ l Modificanon @g g, %, ( h A,,. m m pt .tc Fea 1,21 cm u & I Descriction mo esAm 1a 1c # 9 ~ n.A. l s Oriemator l . Reautrea :Completec ~ Comments Saferv Evaluation i Yes i No Date Imual i 1. Safety Related !M l Y/%l668 l OA Plan 1.2 1.1 Rextor Pool T % Af% l 1.2 Primary Coolant System lXl @% g/f 1.3 Tech Soec Tables 3.1. 3.2 lXl Y/ l/Y/, 2. Safety Analysis )( 4/J 10CFR50.59(a) In 3. SRC Approval ':1 10CFR50.29(c) /r 3f.D 4=le d Tech Spec 6.4 w-l.g lygg %j,.ar-Wa-44 u 'M. l A l l V

.1 Pn.or to implementauon i we, # g,.,,

,.,w # l 3.2 Priortounrestricted g gg gg i ooeration i 4. Tech Soec Chance !b l N 3D 4 mewl w # W l 5. NRC Approval .k 79 4,3 ? 3 10CFR50.59(c) !N,f aPS ,.5;J 0CFR20.1101(b) 6. Alara Review l 'N l ~ /y b* + 6.1 A Heath Physicistinvolved t in ALARA Review? b kl 3g l l 7. Safety Evaluanon Requirements Reviewed l 7.1 Rextor Manager,h'iu M AwDate %jg 7.2 Assistant Manager M( Date Y//Mi O Page 5

AP-201 Fora Suctear Reactor Modificanons i Revision 6:041598 Page 2 of 3 Reamrea l Quality Assumnce Comments Yes i No I 1. QA Plan Arpiicacle h l QA Plan 1.2 2. OA Comnuttee Name i 2.1 Design Engineer yi b,3 h,w 2.2 Quality Engineer I A i Rilu) i ue m l N5 u bb._ - d,Itg tWl,, I 2.3 Ad Hoc i x R i egmrea ; Completed Descnotion Comments .Yes INo Date ilmuall 1. Desien Review lK f2 % d M QA Plan 2.3.1 2. Design Documentation !% l k 8/d QA Plan 4.10.1 cr-w), criss 2.1 Procedure Changes l3 $%g '.B() cr - t!es.p -sa, ar.1 e-o f%-speark. mvs.Jies:A mat , f) i 2.2 VendorManual/ % -++i r. - cu rewwt K I Installation Manual Filed w %+3 - 2swp. 2.3 Drawing Index / QA Plan 10.1.1 ,%u. 22 2.4 System Desenetion lK 176/n adh QA Plan 10.1.2h, su, 2.5 Block Diagram YI QA Plan 10.1.3 2.6 Schematic Diagram l [ QA Plan 10.1.4 2.7 Design Notebook l /g N) QA Plan 10.1.5 3. Procurement Review Nl QA Plan 5. l 5u. me tne:ss(- l1

2h 3Pb QA Plan 6.

G.dev/ %,-A l 4 Matenal Control l su. neu-L(- 4.1 Non-Conforming Material 1 17Mr 3th QA Plan 6.1 G,d,,L Rm,.5L .i. Process Control l Ml QA Plan 7. 6. Insoections $% & a; QA Plan 8. Installation Comoleted "/s a 4 s. Test- !1 !*/<. le 4 o^ ei== 9. 'o'o e-O l l P2,e6 r L

AP-201 Fora Nuclear Reactor Modificanons Revision 6:041598 Page 3 of 3 O' _ Requitec ;Completea l Yes 1.No Date limuali i I i 8.1 5tancara Can~ ranon c y l 8.2 5pecial Test Procecure l7 9. Project File l y QA Plan 2.2.2 l 10. Aucitor Review IXl l l QA Plan 11. j Review Date Completed Date ~ i O l l l I l l { O Page 7 j i

I \\. AP-201 Ford Nuclear Reactor Modifications j Revision 5 I Safetv Evaluation and Crwanlimnee with 10CFR50.59 3 i Is there a change to the facility or procedures as 1 a. described in the safety analysis report? NO,

he safety analysis report specifies a pump with a 20 hp motor. The 20 hp pump / motor being installed is the same pump / motor that was originally installed in 1987 which was replaced on 1996 by a new 25 hp pump / motor l

The 25 hp pump / motor seized and failed March 24 1998 b. Is a test or experiment to be conducted that is not described in the safety analysis report? NO c. Is there a change to the Ford Nuclear Reactor Technical Specifications and/or Safety Analysis? Technical Specifications:

Yes,

icense amendment #44 to be prepared Safety Analysis:

No d. Is there an unreviewed safety questien per 10CFR50.59? YES, An increase in the primary flow rate to greater than 1050 gpm will result in a change in the core delta T allowing an approach to the T.S. Safety Limit of 116 degree inlet temperature. This problem was address by the SRC approved modification # 123 (Title: Install Rundown at 114 'F Tin 3 2 MW). A proposed license change to add a T.S. LSSS = 114 ' F on Tin is currently being submitted to the SRC and the NRC or approval. This unreviewed safety question will be resolved prior to resuming 2 MW Reactor Operations. ALARA Review in Accordance with 10CFR20.1101(b) l a. Are procedures and engineering controls established to achieve occupational doses as low as reasonably achievable?

YES, The pump is located in the reactor basement.

The reactor will be shut down and HP will be present. Page 8

MICHIGAN MEMORIAL PHOENIX PROJECT THE UNIVERSITY OF MICHIGAN E-SAFETY ANALYSIS Ford Nuclear Reactor The University of Michigan Docket 50-2 License R-28 August 1,1993

4. REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 4.1 Primary Coolant System (Figure 4.1)

The components of the primary coolant system are the header and hopper mechanism, the hold up tank, a pump driven by a 20 2 hp electric motor, the heat exchanger, and associated piping and instrumentation. The primary coolant system removes 2 Mw of heat from the core by forced circulation and maintains the bulk pool temperature at less than 116 F. The flow rate is between 900 and 1000 gpm. A movable header is positioned beneath the reactor (~ core to provide forced circulation when the reactor is operated in the forced cooling mode. The header is attached via a rotating flange to primary coolant piping in the pool floor. In the forced circulation mode of operation, primary coolant flows down through the core fuel elements, grid plate, hopper, and into the header. From the header it passes into the holdup tank and to the primary pump where it is pumped to the heat exchanger. From the heat exchanger, primary coolant flows through the primary flow orifice and returns to the pool. Approximately 25 gpm is tapped off downstream of the flow orifice and flows to the primary demineralizer system (hot DI). O

Notes On Maintenance Of Fybroc Pumps O-The following set ofnotes is intended to be used as a guide for disassembly and reassembly of Fybroc pumps. The material contained in this instruction was taken from the Fybroc senice manual and expanded to provide further guidance in areas that have caused problems during previous maintenance. Part numbers refer to the enclosed drawing. Disassemblv:

1. Before disassembly, the electric motor should be either disconnected from its power source. or the switch or circuit breaker must be in an "off" position so that the motor cannot be accidentally staned.

2. Depending upon the fluid being pumped, proper protective equipment should be worn including gloves. safety glasses and necessary radiological equipment.

3. Ensure that the suction and discharge valves are shut and the system drained.

4. If the complete pump is to be removed to a maintenance area, disconnect the suction and discharge piping. Otherwise. leave the pump casing bolted to the inlet and outlet piping.

5. Remove the bolts holding the motor to its mounting plate.

6. Loosen set screws on the me ' mical seal. (Perform this step only sfthe impeller is to be removedfrom the sean ssembly)

7. Remove the casing bolts (lD), nuts (1C), and washers (1E) holding pump adapter (71) to the pump casing (1). Remove the shims (67) from each side of the pump.

(Be sure to note the number ofshims removed)

8. Pull the motor and impeller assembly away from the pump casing and remove the cover o-ring C3h The pump is now ready for inspection and/or repair.

9. To disassemble the impeller assembly, remove the locking ring (14B) by removing the two allen head screws. Remove the segment key (14A).

10. Remove the rear fan cover from the motor. Using vise grips or a pipe wrench to f

hold the rear motor shaft extension from turning. (Do nor hold thefan as this

will result in damage to thefan assembly) Remove the impeller assembly by I rotating the impeiler (2) in a counter-clockwise direction as viewed while facing O_ the impeller. It may be necessary to use a strap wrench or similar device to loosen the impeller but extreme care must be used to prevent damage to the impeller. Once the impeller has been unscrewed, the impeller and seal assembly can be removed. I 1. If required. the impeller (2) can be removed from the seal assembly by pulling on the seal assembly with a twisting motion.

12. If required. The gland ring (17) can now be removed by disengaging the four gland bolts 17C) securing the gland ring to the cover (11), be sure to note the i

arrangement of gaskets or o-rings for reassembly. The carbon and ceramic elements of the seal should be handled carefully to prevent, chipping, scratching, or contact with skin oil.

13. The motor shaft adapter (6) may be removed from the motor shaft by temoving the allen head capscrew (6A). The drive key (46) will slide off with the adapter.

In the bore of the shaft adapter may be a series of washer like shims (6B), it is important to note the number of shims that are in the bore for reassembly. Reassembiv: ()_

1. Ifit is necessary to check the direction ofrotation of the motor, verify direction of rotation before begmning assembiv of the pump. (Past crperience has shown that testing direction ofroucieu ofthe pump with the impeller installed can result in binding or damage to the impeller.)

2. Steps 3 to 6 are for installation of the mechanical seal. If the mechanical seal was not disassembled then skip to step 7.

3. Wear gloves when handling seal materials to prevent contammation with skin oil.

Install the seal carbon insert with gasket in the rear of the stuffing box cover. Place the gland ring with the second gasket over the insert. Make sure all gaskets and gland pilots have been properly installed and engaged. Install seal guard and gland bolts i 17C) and tighten the bolts evenly, cross-staggering tightening of gland bolts. Excessive gland bolt pressure can result in distortion of the stationary insen. Gland bolts should be torqued to a maximum of10ft-lbs.

4. Coat the impeller sleeve with oil before installing any seal parts. Use low i

viscosity oil. SAE-10 or equivalent. Install the impeller sleeve through the l stuffing box cover. being careful not to chip the stationary insert. O

l

5. Engage the compression unit containing the collar, springs, and compression ring i

with the seal ring. Be sure the drive pins in the compression ring engage the slots Q in the seal ring. This is the complete rotating assembly. l

6. Install the rotating assembly over the impeller sleeve. Use a slight twisting motion. Place the rotating assembly up against the carbon insert ring.

l

7. Install the shaft sleeve (6) with the drive key (46) and shims (6B) over the end of l

the motor shaft and secure with allen head cap screw (6A).

8. Using a pair of vise grips or a pipe wrench to hold the rear motor shaft extension j

from turning, placa the impeller assembly over the end of the motor shaft and turn j impeller in a clor ovise dhection. when facing the impeller, to attach the impeller to the shaft. Ensu that 6 impeller is fully engaged on the motor shaft sleeve j .(6).

9. Install the segment key (14A) and secure in place using locking ring (14B) and -

two allen head screws.

10. Coat the cover o-ring (73) with a silicone lubricant and position in cover (11).

I1. Install the casing shims (67). I Caution: If the casing (1), impeller (2), shaft sleeve (6), cover (11), pump or motor adapter (71 or 71B respectively), or the motor itselfis replaced, the proper number of shims (67) will have to be deternuned as outlined in Note 1 of the general notes section of this procedure. t

12. Install the motor and impeller assembly into the pump casing and secure in place j

using casing bolts (ID), casing washers (IE), and casing nuts (1C). Tighten j casmg bolts to 10ft-lbs using a proper torque sequence. I

13. Install rear fan cover for motor and secure in place.

14. If the set screws on the mechanical seal were loosened, follow this step, otherwise skip to step 15. Adjust the spring gap dimension to the value stamped on the seal I

collar. Tighten the set screw and make sure the spring gap is equally spaced. l i

15. If the pump casing was disassembled from pump inlet and outlet piping, reposition pump and reattach piping. For 4" flanges the torque spec is 27-36 ft-lbs. For 6" flanges the torque spec is 35-50 ft-Ibs.

16. Bolt motor to the motor support structure.

17. Reconnect electrical power to the motor ifit was removed.

O O w

18. Clear tags if necessary and open supply raa discharge valves. When system is filled and vented. perform an operatior.al check of the pump. Speciscally verify (y

proper direction of rotation of the pump, listen for abnormal noises in the pump, v and check for overheating of the motor. f 1

19. When the pump is retumed to normal service. periodically monitor motor temperature and pump flow output in order to assess the pump for continued proper operation.

l l General Notes: l Note 1: When required to adjust the clearance between the cover (11) and the i impeller (2) or between the impeller face and the pump casing, shims (67) will have to be added or removed per the steps below.

1. Loosen the allen head screws on the mechanical seal. This prevents seal face damage during adjustment of the impeller clearance.

2. Remove a shim (67) and hold the cover (11) against the pump adapter (71). With the cover nrmly held against the pump adapter, measure the clearance between the impeller and the cover. The desired clearance is.020". Add or remove shims (67) as required to obtain the desired clearance.

,O

3. Continue with the reassembly procedure. Note that by adjusting the l

casing to impelle.r clearance the impeller face to the pump casing clearance will also be arTected. To ensure proper impeller face to l pump casing clearance during assembly, rotate the rear motor extension by hand while tightening casing bolts (ID) to detect for binding. Should binding occur additional shims (67) will have to be added.

4. Repeat the above process until all clearances are properly established.

l I O

Lessons Learned From The Past: -Rotation check with pump completely assembled resulted in binding of the pump due to the impeller spinning off because of wrong rotational direction. -Rotation check with pump impeller assembly attached resulted in the impeller unscrewing and binding on the shaft sleeve due to incorrect rotational direction. -Installation of new motor resulted in incorrect impeller to cover clearances due to variance in the length of motor shaft. O.. 4 l l O

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Memo To: modification request file #129 From: Christopher Brannon (Elec Eng lil)

Subject:

Phone conversation pertaining to 25 hp motor failure and motor grease channel defect 1 4/7/98 I spoke to John Keinz head of Quality Control at US Electrical Motors John stated that a manufactured defect would be covered by their warrant process and ask the motor be sent to a Michigan local Electrical Repair Facility and the Facility contact him with an inspection report. 4/7/98 Spoke to John Espy at Fife Electrical in Detroit and set up a delivery date and time and inspect. 4/9/98 Eric Sharp delivered motor to Detroit and received a warrant receipt from Fife Electrical iO 4/13/98 John Espy at Fife Electrical call me and confirmed that the manufactured defect would be covered by warranty and the US Electrical Motors would replace the motor at no cost. Fife Electric will contact me for motor pick up. LO

(. TELECON

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1 i TO: Modification Request # 129 FROM: Bemard Ducamp 3._ I) ^ Ass't Mgr, Rx Operations T DATE: April 7,1998

SUBJECT:

Motor Greasing 1 I had a conversation vM Lee Miller, a supervisor in the U of M's Machinery Repair Group regarding the zire fitting (where grease is injected) and drain plug (where old g ease is ejected) locations on an electric motor which was rebuilt for the U of M by Arrow Motor and Pump (Job # R7803). I told Lee that we had received this electric motor with the zire fitting located below the drair, plug. I considered this abnormal since in most lubrication setups, the fresh satff goes in from above and the old stuffis removed, with gravity's assistance, from below. 'O V Lee said that the zire fitting and drain plug could be exchanged. I asked if he wanted to consult with the vendor or rebuilder. Lee said that was unnecessary. He said that the answer was obvious to him from his many decades of experience in machinery repair. I concur with him, since the rebuilt motor has never been run, so the bearing cavity, fill and drain lines are all full of fresh grease, with no old stuff to gum things up. 4 u.s mete, as up ErtL*4 239 O M Wuf Tyr Jf t -T r. ~ J'Mc3 e...a %, 10. LJ Tqge1 of1 l 1

f,. StLSERSEBES 2nqES I O v Modification Request #129 i i I l Reason for superseding: ) SAFETY REVIEW COMMITTEE MEETING MINUTES Agenda-First SRC Spring 1998 Meeting FNR Operation in Forced Convection. Safety Evaluation for a Proposed License Change Building: Phoenix Lab Room: FNR 3102 Date: April 1,1998 Time: 15:00 1 "The SRC requested that FNR personnel re-perform page 2 of the modification request checklist, /, 3aying careful attention to each item. The SRC was not concemed with the adequacy of the work, )ut rather with the thoroughness of the documentation. The SRC asked to see this page 2 again after FNR staff's review. The SRC also requested that the associated safety evaluation be rewritten to indicate that a change to Tech Specs is required: the proposed license change (Ref. 9)" { I !Os { l ) 1

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i i THE LS.nnstrr OF MIGUCAN 9+ CourcrOFEsco.muNG l wctua e,canec no uoxxacnt mcu ! ( mootn etum Ns*m" cut wtScS4Emm 313 M42e0 FAX 313 WMO Memorandum to: F rderick C. Neidhardt Acting Vice President for Research d b[h From: John C. Lee. Chair Ford Nuclear React. Safety Review Committee

Subject:

Annual Report on FNR Safety and Operations Date: August 26,1997 The Safety Review Committee (SRC) for the Ford Nuclear Reactor (FNR) submits its second annual report summanzmg the operational and safety aspects of the reactor facility over the period 1 July 1996 to 30 June 1997. The SRC met three times during this reporting period to review safety-related matters and to approve proposed facility modifications. In preparing this repon, the committee reviewed the'FNR Report of Reactor Operations for 1996 (Attachment), SRC meeting minutes and FNR Cycle Summaries, and obtained assistance from FNR management. We are pleased to repon that the FNR operated safely during the reporting period, as evidenced by no violations cited as part of'the annual U. S. Nuclear Regulatory U) Commission (NRC) inspection and experiencing no Reportable Occurrences. Furthermore, an external audit and emergency drill indicate the facility is well maintained, with proper management attention to safety. A number of facility modifications and upgrades were made, including the replacement of a safety-channel neutron detector. Unscheduled Shutdowns During the reponing period, there were 16 unscheduled shutdowns of the reactor, compared with an average of 19 per year over the past 10 years. Eight of the shutdowns were due to electronic noise in the reactor safety system. Of the remainmg eight, (a) three were due to electrical power outages, (b) two were caused by secondary cooling pump problems, (c) two were due to reactor system and control circuitry problems, and (d) one was due to health problems of a lead operator. One shutdown occurred during an ice storm, when a power flicker dropped, without tripping the primary pump, a movale header that directs the primary flow. Subsequent effon to restart the reactor resulted in a High Power / Header Down scram. The NRC concurred that this unanticipated event was not a Reponable Occurrence. A " Header Up" indication was installed on the control room console to avoid a repetition of the incident. Following extensive troubleshootmg effons over the past several years, one 20-year-old neutron detector in Safety Channel A was replaced with a new ion chamber in January 1997. Since then, the FNR has experienced no shutdowns due to electronic noise, which suggests that one major cause for unscheduled shutdowns has been eliminated. We note i that none of the unscheduled shutdown events, including the High Power / Header Down scram, entailed unsafe operating conditions for the FNR.

2 Facility Modifications and Repairs

1. Facility Modifications In addition to the installation of a " Header Up" indication discussed earlier, the SRC approved modifications in the temperature recorder rundown circuitry to use the core inlet temperature directly as the scram setpoint, rather than inferring the inlet temperature from outlet temperature measurements. This modification will provide clear compliance with the safety limits specified in the FNR license.

2. Repairs and Replacement In addition to the safety-channel ion chamber discussed earlier, the secondary cooling pump motor was replaced. Problems with the old motor caused two unscheduled shutdowns, as noted on page 1.

3. Modifications under Consideration i

Study is underway to design and install an above-water pipe for the pneumatic tube system. The design will be evaluated following an ALARA (as low as reasonably achievable) review j on radiation exposure due to the new piping arrangement. Experiments and Tests

1. Mercuric Oxide (HgO) Irradiations t

i An experiment involving the irradiation of HgO was approved by the SRC, after a trial irradiation with surrogate materials. Initial irradiations of HgO performed satisfactorily and I the encapsulation systems remained intact. Further irradiations of HgO at the FNR will require periodic SRC review and approval.

2. Heavy Section Steel Irradiation (HSSI) Program

'Ihis multi-year project investigates the neutron irradiation embrittlement of metal alloys and is funded by the NRC, through.0ak Ridge National Laboratory. The HSSI project will provide critical data for the long-term integrity of reactor pressure vessels, which plays an 3 important role in the plant life extension of existing nuclear power plants. The project 1 involves the participation by the FNR. the Department of Nuclear Engineering and Radiological Sciences, Oak Ridge, and the University of California at Santa Barbara. Radiation Protection Issues

1. Radiation Protection Program and Health Physics Procedure Audit The annual Radiation Protection Program audit was conducted by the Radiation Safety Officer, Mark Driscoll,in December 1996. Numerous discrepancies were notsiin areas such as labeling, posting, material storage, and procedures. An extensive effort was initiated to review and update all of the FNR Health Physics procedures, with Health Physics Procedure No.116 for Co-60 irradiator checks serving as a template. The SRC approved a revised version of HP-116, together with suggestions for editorial and structural improvements for HP procedures in general.

2. Neutron Beam Port Exposure Incident A researcher was inadvertently exposed to the neutron beam at J-port in October 1996. The exposure was initially estimated to be as high as 38 mrem, although the actual exposure was

l' 3 determined to be <10 mrem, which is small compared with an occupational exposure limit of p_ 5 rem / year. The beam port has been shut down while facility modifications are being made Q to provide a positive indication of the beam port status and to preclude access to the beam port area when a high neutron field is present. Radiation incidents of this type, although not considered to impact the health and safety of the public, will also be reported to the SRC. Facility Inspection Audit, and Emergency Drill 1.NRC Inspection Routine inspection of the FNR was conducted by Timothy D. Reidinger and Thomas M. Burdick in Febmary 1997. No violations or deviations were identified during the inspection l j and the overall FNR operation was characterized as first class, comparable to that of the 10-MW University of Missouri Research Reactor. The inspectors, however, recommended l l enhanced oversight of the health physics aspects of the FNR operation, together with closer coordination between the health physics staff and FNR management.

2. Reactor Audit Ward L. Rigot, Reactor Supervisor, Dow TRIGA Reactor, performed the annual audit of the facility in November 19%. One specific suggestion from the audit resulted in the modification of the temperature recorder rundown circuitry discussed earlier. Some wording l

changes to the control rod calibration procedures were suggested and have been duly implemented. Wording changes suggested for the Quality Assurance Plan will be incorporated, together with more substantial changes,into the next submittal to the NRC.

3. Emergency Plan Meeting and Drill O

ne annuai emerSencx driii. cem8ined with a review ef the emeriencv 9 an. was successfulix i conducted at the FNR in May 1997. The drill included representatives from the Ann Arbor Fire Department, UM Department of Safety, and UM Radiation Safety Service. Editorial changes to the emergency procedure are being made as a result of the drill. In addition, effort is underway to follow up the recommendation that the meteorological tower at the Space Research Building be utilized to monitor the meteorological pattern around the FNR. FNR Management and Personnel Be FNR has operated since September 1 through a team approach, with Bernard Ducamp, assistant reactor manager for reactor operations, serving as interim reactor manager, and with active management collaborations by other assistant managers and the electronics engineer, The Phoenix Project director has also increased his involvement in day-to-day affairs of the reactor. The vacancy in the reactor manager's position has, however, resulted in delaying shipments of spent fuel and a few facility modifications. The FNR has hired five operations staff members to replace six licensed staff members who left during the reporting period. Four of the new staff have acquired either the Reactor Operator or Senior Reactor Operator license. The FNR had to adjust its operating schedule during the summer of 1996, but returned to its normal schedule in September. . Attachment xc: R. F. Fleming (w/o attachment) Safety Review Committee (w/o attachment) Phoenix Executive Committee (w/o attachment) FNR Operations Staff (w/o attachment) O I

1 [ MEMORANDDI [ G( w TO: Safety Review Committee Members: l Professor Dale Briggs Professor I-Wei Chen Professor Henry Griffin Professor John King Professor John Lee ) l Professor James Martin 1 Professor Fredrick Neidhardt Mr. Mark Driscoll Mr. Douglas Wood Professor Ronald Fleming Director Philip Simpson. Assistant Manager FROM: Bemard P Ducamp t Assistant Manager. Reactor Operanons DATE: April 21.1997

SUBJECT:

External Auditor's Report November 1996 A Attached is the audit report from Mr. Ward Rigot. Reactor Supervisor. DOW TRIGA Reactor for Ford Nuclear Reactor. This audit satisfies FNR Tech Specs section 6.2, " Review and Audit." The responses to the recommendations are provided for your consideration. Recommendation 1: "It is recommended that the TS relative to the safety limit and limiting safety system settings be reviewed for appropriateness." Response 1: Reviews will begin this Spring at managers' meetings to address this recommendation. The issues involved are fairly complicated have a long regulatory history. and merit careful review. Inputs or participation from SRC members is welcome. Modification Request #123. " Install temperature recorder rundown at com inlet temperature of i 16 F when at 2 MW." has been initiated to permanently address the existing Tech Spec Safety Limit wording. Until this Modification Request is implemented, a Temporarv Operating Instruction has been issued which reduces the Temperature Recorder rundown serpoint from 129 F to 126 F. l [ Recommendation 2: "The procedure for the control rod calibration should be revised to l include either the procedure necessary to complete calculations of control rod worth; or, point to the location for instrucuons to complete these calculations: or. describe attemate ~ ?- Tage1 of2

methods for calcuiation of controi roc n ortn. such as graph paper and french curve methodology. A Resnonse 2: Roa calibrations for FNR are done using Calibration and.\\laintenance Procedures CP 201 iShim-safety rod calibrations: and CP-202 (Controi rod calibration). CP-201 was revised in the followmg ways. as The oeriod cer roa ruil o mcreased from 45 to about 60 seconds to avoid 30 second dynamie reaks a kn/dt). The 30-second control rod inhibit from the period recorder.'when actuated. provided undesirable distraction and intermption in the calibration procedure. b) The plotting ponion of the procedure now specifies the objectives without mandating a method. c) Table 1. "Informanon." was revised to document the methods and tools used for plotting rod worth curves. Table i now includes a signature from the Shift Supervisor. as well as a management review, to verify that the fit between the curves tpiotsi anc the actual calibration data was checked and is acceptable. CP-202 was revised in the following ways: a) Control rod calibration is now done in one continuous sweep rather than the previous two overlapping sweeps. A single sweep provides better continuity of data. The individual pulls are small enough to easily avoid any 30 second dynamic peaks adiSKVdt). l b) The plotting portion of the procedure now specifies the objectives without mandating a method. c) Table 1. "Information and Data." was revised to document the methods and tools used for plotting rod wonh curves. Table 1 now includes a signature from the Shift Supervisor, as well as a management review. to verify that the l fit between the curve iplot) and the actual calibration data was checked and is acceptable. Recommendation 3. It is recommended that the " Quality Assurance Plan" be edited tc clarify the intent of the annual review. t l Response 3: This is a quick editorial change. but QA Plan changes require NRC approval. Two possible approaches are i 1) make the change and contact NRC for approval, or (2) prepare a memo to the appropriate files which clarifies the referenced point without revising the QA Plan wording. O f.:qe2 of2

1 Maana.Wngan 48667 January 16,1997 i l Dr. Ronald Fleming Director, Phoenix MemorialProject l The University ofMichigan 2301 BonisteelBlvd. Ann Arbor, Michigan 48109 Dr. Fleming; Q Enclosed is my report covering the audit of The Ford Nuclear Reactor performed on November 11-13,1996. This audit was a review of the program which is required by technical specifications. My overalliwdons were favorable. There were no noted areas of de5ciencies. Re:+ m- =dadons are for you and your staffs consideration and do not reflect any l unsafe condition. i n Ward L Rigot i Reactor Supervisor l Dow TRIGA Reactor l l cc: B. P. Ducamp Assistant Reactor Manager Ford Nuclear Reactor O

SUMMARY

l An audit of The Ford Nuclear Reactor was performed November 11,12 and 13,1996. l This is a routine annual peer review audit as specified in Technical Specification 6.2.9.b. l The facility appears to be maintained in a safe condition. No areas of non compliance l were observed. There appears to be strong management commitment to continuing safe operations. The following areas were covered during this audit A) Review of the Technical Specifications B) Review of Operations c) Review of Calibration and Maintenance Activities D) Review of Facility Modifications l E) Review ofThe Quality Assurance Plan F) Review of The Transition Management G) Review of Training and Qualification H) Review of The Safety Review Committee Minutes These reviews were condumi through review of facility h- = :an, intarviews with management, o[maasuis and support staff and through observation of operations. l The following runmmenderinne are intended for review by facility management for l Consideration. Recommendation 1 : It is recommended that The TS's relative to the safety limit and limiting safety system seuings be reviewed for ayrvr:ar-? Recommendarian 2 : The procedure for the control rod calibration should be revised to include either the procedure n-==ry to complete calculanons of connel rod worth; or, point to the location for instructions to complete these calculations; or, describe e-:-:+;= Ale alternate methods for calculation of control rod worth, such as graph paper and l-french curve methodology. Recommendation 3 : It is recommended that the ' Quality Assurance Plan' be edited to l clarify the intent of the annual review. O f

l A. Review of Technical Specificanons f The Technical Specifications (TS) were reviewed and discussions held initially with the assistant reactor manager (ARM) and subsequently with the facility director. After review of the TS. it seemed that an unusual set of conditions were defmed as the safety limit for the facility. TS 2.1.1 ' Safety Limits in the Forced Convection Mode' sets the j i limits as l

1. The true value of reactor power t P) shall not exceed 4.68 Mw and the true value i

of flow (M) shall not be less than 900 gpm.

2. The true value of reactor coolant inlet temperature (T ) at 2 Mw shall not exceed i

116 7. l

3. The true value of water height above the core (H) shall not be less than 18 feet while the reactoris operating.

Facility management explained that these conditions replaced a graphical repreentation l (a family of curves) of conditions which will assure that the peak temperature in the fuel claddag ch2aml will be below the boiling point of water in those chaaaa3= And although facility management preferred the family of curves concept negonanons durmg the license amendment process yielded this safety limit. The safety limit became imposi.ut after review of temporary operanng mstructions and Q facility modifications. A change in the prunary coolant pump iwavmi pd-sy coolant flow by appronmately 10 percent. And although this viewed as a benefit to safe operations, it apparently has caused a problem in ma ntaining the core inlet tvano below the safety limit with the reactor scram of 129 T on the core outlet. Although the scram was adjusted to lower the scram setpoint, it ratsed the question as to why the core outlet temperature was used as the scram setpoint with inferences made with regard to l core delta T as a measure of reactor power. If core inlet 44uss is the safety limit the i scram should be placed on the core inlet temperature. R-===d=*ia= 1 : It is recommended that the TS relative to the safety limit and limiting safety system semngs be reviewed for appropriateness. B. Review of Operations l The following evolutions were observed.

1) Fuel Movement. The procedure was conducted to implement wa==y core configuration and as a trauung exercise for operator tramees. In the connel room was an SRO (lead) and two trainees. The procedure was available and referenced during the evolution. Review of the procedure was dien=ad prior to the evolution. At the reactor bridge,2 SRO's and one trainee performed the actual fuel movement after instructions from the control room were received. Excellent communication was

observed. Confirmanon of all operanons was made prior to implementation. All manipulations were adequate and performed within the requirements of the procedure. i V There was some difficulty in removing the fuel tool from the sample folder during reinsertion of the sample holder into the core grid. This poses no safety coccern and improvements are certam to be made with additional manipulations.

2) Reactor Startup. The initial reactor startup and core excess measurement was observed. The procedure was avadable to the operator and trainees. All 1

manipulations were adequate and performed within the requirements of the procedure.

3) Control Rod Calibration. A portion of the control rod calibration was observed..

The procedure was available and referenced during the evolution. Following completion of reactivity manipulations relative to the procedure; the instructions for j completion of the computer generated calculations were not available to the operations staff on shift. All necessary informanna needs to be availabic to the operators to complete this procedure. Reenmmendatinn 2 : The procedure for the control rod calibration should be revised to include either the procedure necessary to complete calculations of control rod worth; or, point to the location for instructions to complete these calculations; or. describe j acceptable alternate methods for calculation of control rod worth such as graph paper and i French curve methodology. O C. Review ef Calibration and xaintenance Activities Several calibration wotMees were reviewed and found to be adequate. There were no unplanned maintenance activities for review. D. Review of Facility Modifications i 1 l Facility madificMons as described in the Quality Assurance Plan were reviewed. Replacement of the prunary coolant pump improved coolant flow by appronmaraly 10% and will prove to be a benefit to operations. It did however raise a concern relative to a need to modify the method by which safety limits are assured (see recammendarino 1). E. Review of The Quality Assurance Plan The Quality Assurance Plan was reviewed in detail. The plan defines the methods of review and dacamentation necessary for facility modifications and fuel shipments. It is a very detailed and complete documentation. There is however a need to clanfy the intent on one section of the plan. Section 2.4 of the plan states that The Quality Assurance Plan will be reviewed annually by a consultant appointed by the Safety Review Committee (SRC). Upon discussion with facility management it was determined that the intent is for the extemal audit review the facility modifications and fuel shipments described in the h plan as opposed to the planitself. L

p Recommendation 3 : It is recommended that the ' Quality Assurance Plan' be edited to (/ clarify the intent of the annual review. F. Review of The Transition Management The level of management for the facility has reduced through the retirement of The Reactor Manager (RM). My understanding is that a review of the long term future of the facility is in pmgress and upon completion a search for a replacement will begin. During this transition period the Assistant Reactor Manager (ARM) is fulfdling the responsibilities of the RM until the search is complete. The facility has hired a consultant familiar with the facility to assist in this area. The consultant is the former ARM and served in that role for many years prior to his retirement. Although the facility leadership appears dedicated and committed to safe operations, I would encourage filling the RM role in a timely manner. G. Review of Trainmg and Qualification There have been four new reactor operator trainees on staff since the previous audit. I met with three of the trainees and discus =d heir qualifications and reviewed the traming t pmgram which qualified them for shift operations. The qualibdons and trammg is adequare to assure knowledge of operations and procedures. Although a burden is placed O of the ' Lead Operators' during these times, as far as restriction to the control room, the benefits to the trainees is essential for their experience. H. Review of The Safety Review Committee Minutes The minutes of the previous two Safety Review Committee Minutes were reviewed. They appear to be complete and document the actions of the committee. At tunes there was merely a sentemear that an item was di= ens-A with no detail When asked, facility management provided sufficient detail for understanding. I would encourage that, where appropriate, sufficient detail be included in the minutes to serve as a stand alone document. O

~ L MEMORANDUM TO: Licensed RO's and SRO's FROM: Bernard Ducamp Ass't Manager, Rx Ops DATE: February 27,1998

SUBJECT:

Summary Report ' NRCInspection of FNR Operations l Here is a brief summary report on this past week's NRC inspection. i FNR operations was generally found to be quite satisfactory (read as "A-o.k."). No deficiencies or problems were found in the performance of and documenting j of any reactor evolutions, maintenance, or repairs. Licensed operators and leads are obviously doing a very good job. I thank you very much for that. i!O One minor area that was discussed by the inspector involved the documenting of unplanned or corrective maintenance. Whether or not we modify any procedures to improve our documentation of unplanned and corrective l. maintenance will depend on the final inspection report. l One major area of discussion related to the primary coolant system modification (removing the check valve internals and changing out the pump) and its impact on core delta-T. The subsequent modification to install a T-inlet =116 F @ >80% power rundown was initiated about 6 months later. The NRC inspector sees this as an open item - should we have realized the impact of a core delta-T reduction L on the vulnerability of one of our safety limits in April rather than 6 months . later? The regulatory hammer is 10 CFR 50.59, " Changes Tests and Experiments." The NRC inspector was also concerned about whether everyone appreciated the significance of violating a safety limit. I am routing a copy of the rule 10 CFR 50.36, which states what must be done if a safety limit is exceeded. . In addition, the installation of the primary pump (visqueen and duct tape over the motor terminations, and banding material in place of motor mounts) was

discussed by the inspector. We could not defend these items as being satisfactory, so I committed us to putting a permanent metal cover on the motor

~ . terminations, and to installing proper motor mounts. T4qe1 of2

( l l l Emergency preparedness and health physics will be covered in a future l mspection. The NRC is trying to get its non-power reactor inspectors to visit more often. l That means more detailed and in-depth inspections in the future. If there l continues to be a constructive and healthy relationship between us and the NRC l l (it's currently verv good) and we have a quality inspector (Tom Burdick is very l professional and reasonable), then we have no room to complain. Please feel free to ask questions on the inspection, or to offer suggestions on how we can improve our work and productivity. ' (~T %) l l l 1 O O Tage2 of2

~. ,( a h Nuclear Regtdatory Commission

tintd, 3g.g the license and the regulations in this
earch nably and submit a Licensee Event Report to chapter.

the Commission as required by 150.73. (27 FR 12n$. Dec. 3,1962, as amended at 31 4"I8' Licensees in these cases shall retain FR 12780, Sept. 30,1966; 35 FR 5318, Mar. 31, res or the records of the review for a period of basis 1970; 35 FR 6644, Apr. 25, IM. 35 FR 11461. July 7,19701 three years following issuance of a Li- >le as-censee Event Report. sticus 650.38 Technical specifications. (B) Bafety limits for fuel reprocess-or be-appli-(a) Each applicant for a license au-ing plants are those bounds within 1ctitu thorizing operation of a production or which the process variables must be aking utilization facility shall include in his maintained for adequate control of the application proposed technical speci-operation and that must not be ex-a con-r,th2 fications in acconiance with the re-ceeded in order to protect the integrity quirements of this section. A summary of the physical system that is designed ucted Misu statement of the bases or reasons for to guard against the uncontrolled re-I h and such specifications, other than those lease or radioactivity. If any safety covering nAminiatrative controls, shall limit for a ihel reprocamming plant is also be included in the application, but exceeded, corrective action must.be shall not become part of the technical taken as stated in the technical speci-

4 181~

specifications. fication or the affected part of the '("$ (b) Each license authorizing oper-process, or the entire process if re-ndings ation of a production or utilization fa-quired, must be shut down, unless this cility of a type described in 150.21 or action would further reduce the margin modi. ( 650.22 will include technical specifica-of safety. The licensee shall notify the tions. The technical specifications will Commi== ion, review the matter, and 1 3 L be derived from the analyses and eval-record the results of the review, includ-S ing the cause of the condition and the f uation included in the safety analysis n t basis for correctiv 3 "P' report, and amendments thereto, sub-preclude recurrence.e action taken to y a fea" mitted pursuant to $50.34. The Com-If a portion of the 3PDll-mission may include such additional process or the entire process has been technical specifications as the Com-shutdown, operation must not be re-1 cp-f (ncor-mission finds appropriate. sumed until authorized by the Commin. [.

cant, (c) Technical specifications will in-sion. The licensee shall retain the clude items in the following categories:

record of the results of each review i prov-from (1) Safety limits, limiting safety system nutil the Commission terminates the settings, and limiting control settings. license for the plant. sc:n-ission (1)(A) Bafety limitefor nuclear reactors (ii)(A) isimiting safety system set- .te in are limits upon important process vari-tings for nuclear reactors are settings ables that are found to be necessary to for automatio protective devices relat-as re-peri-reasonably protect the integrity of cer-ed to those variables having significant !sults tain of the physical barriers that guard safety functions. Where a limiting safe-against the uncontrolled release of ra-ty system setting is specified for a rams dioactivity. If any safety limit is ex-variable on which a safety limit has 111 be ceeded. the reactor must be shut down. been placed, the setting must be so is, h licennee shall notify the Commis-choson that automatic protective ac-cility sion, review the matter, and record the tion will Correct the abnormal situa-c nse tion before a safety limit is exceeded, remnits sairn of the review. including the

ed to cause of the condition and the basis foF If, during operation, it is determined o the corrective action taken to preclude r6 that the automatic safety system does af;ty currence. Operation must not be re-not function as required, the licensee sumed until wthorized by the commis-shall take appropriate action, which may sion. The ' licensee shall may include shutting down the reactor.

] retain the tim 3 record of the results of each review The licensee shall notify the Commis-1(* until the Commission terminates the sion, review the matter, and record the ealth license for the reactor, except for nu-results of the review, including the v.m cause of the condition and the basis for clear power reactors licensed under corrective action taken to preclude re-e en-150.21(b) or $50.22 of this part. For ty in ts of these reactors, the licensee shall notify currence. The licensee shall retain the record of the results of each review the Commission as required by 150.72 until the Commission terminates the

license for the reactor except for nu-a nuclear reactor not licensed under clear power reactors licensed under $50.21(b) or 550.22 of this part or fuel re-550.21(b) or 650.22 of this part. For processms plant, the licensee shall no-these reactors the licensee shall notify tify the Commission review the mat-the Commission as require'i by 550.72 ter, and record the results of the re-and submit a Licensee Event Report to view. including the cause of the condi-the Commission as required by 550.73. tion and the basis for corrective action Licensees in these cases shall retain taken to preclude recurrence. The 11-the records of the review for a period of censee shall retain the record of the re-three years following issuance of a Li-suits of each review until the Commis-T s censee Event Report. sion terminates the license for the nu-(B) Limiting control settings for fuel clear reactor or the fuel reprocessing reprocessing plants are settings for plant. In the case of nuclear power re-automatic alarm or protective devices actors licensed under $ 50.21(b) or $ 50.22. related to those variables having sig-the licensee shall notify the Commis-niacant safety functions. Where a lim-sion if required by $50.72 and shall sub-iting control setting is specifled for a mit a Licensee Event Report to the variable on which a safety limit has Commission as required by 550.73. In i been placed, the setting must be so this case, licensees shall retain records chosen that protective action, either associated with preparation of a Li-automatic or manual, will correct the censee Event Report for a period of abnormal situation before a safety three years following issuance of the limit is exceeded. If, during operation, report. For events which do not require the automatic alarm or protective de-a Licensee Event Report, the licensee vices do not function as required, the shall retain each record as required by licensee shall take appropriate action the technical specifications. to n sintain the variables within the (11) A technical specification limiting limiting control-eetting values and to condition for operation of a nuclear re-repair promptly the automatic devices actor must be established for each item or to shut down the affected part of the meeting one or more of the following process and, if required, to shut down criteria: the entire process for repair of auto-(A) Criterion L Installed instrumenta-matic devices. The licensee shall notify tion that is used to detect, and indicate K the Comminston, review the matter, in the control room, a significant ab-and record the results of the review, in-normal degradation of the reactor cool-cluding the cause of the condition and ant pressure boundary. the bacia for corrective action taken to (B) Criterion 2. A process variable, de-preclude recurrence. The licensee shall sign feature, or operating restriction retain the record of the results of each that is an initial condition of a design review until the Comminaion termi-basis accident or transient analysis nates the license for the plant. that either assumes the failure of or (2) Limiting conditions for operation. (1) presents a challenge to the integrity of Limiting conditions for operation are a fission product barrier. the lowest functional capability or per-(C) Criterion J. A structure, system, formance levels of equipment required or component that is part of the pri-for safe operation of the facility. When mary success path and which functions j a limiting condition for operation of a or actuates to mitigate a design basis i J nuclear reactor is not met, the licensee accident or transient that either as-i shall shut down the reactor or follow sumes the failure of or presents a chal-I any remedial action permitted by the lenge to the integrity of a fission prod-1 I technical specifications until the con-uct barrier. dition can be met. When a limiting (D) Criterion 4. A structure, system, condition for operation of any process or component which operating experi-step in the system of a fuel reprocess-ence or probabilistic risk assessment ing plant is not met, the licensee shall has shown to be signiccant to public shut down that part of the operation or health and safety. follow any remedial action permitted (iii) A licensee is not required to pro-by the technical specifications until pose to modify technical specifications the condition can be met. In the case of that are included in any license issued i 648 r

NUCLEAR REACTOR LABORATORY l l The Attached Correspondence is Routed To: iO jV ADMINISTRATIVE REACTOR OPERATORS i t N lackbum. R. M B $ Berg C. I Corev. D. Y Hall. T. T#f Ducamo. B. 99 y' Huck.B. Dunbar. W. ( Landis. M.

dd/

l Flemina R. F X uoCue.R. 3 h Sharp. E. Ketola. Z. Lum. H. Mullins. M. k X Simoson. P. ~ PROFESSIONAL / TECHNICAL HEALTHPHYSICS X Brannon. C. Downey,H. h Hartman. M. McCall. W. D Lindsay. J. [M l X Nowak.S. X Touchberry. E. Y NUCLEAR MEDICHE [./ k Weaer. C. Hadley. J. Hagen,C. For Your Action Per Our Conversation For Your Approval Per Your Requen I For Your Comments Read and Pass On b For Your Information Please Retum To: For Your Review p)u fy l For Your Sionature l REMARKS 7)v T. M a u u FM2 S a. k ' L2 m Ys. " uJ \\ ~.) t/s.?,/ ge 3 /,O DATE: ORIGINATOR:

viCHIGAN MEMORIAL-PHOENIX PROJECT i PHOENIX MEMORIAL LABORATORY FORD NUCLEAR REACTOR ANN ARBOR. MICHIGAN 48109-2100 I am<e a tw oma I To: John Lee, Chair l Safety Review Committee - FNR @ [' From: Ron Fleming, Director j A MM-PP Date: March 23,1998 About: Tech Spec Definitions of " Safety Limits" and " Limiting Safety System Settings" In your email to Bernard Ducamp of March 17th, you wisdy suggested that we provide the members of the SRC with a clear distir.ction of the meanings of the terms " Safety Limit" and " Limiting Saicty System O settino". es they are used in our Tech specs. This *iii be per icuieriv l i valuable for our new member. As a first cut at this I am including with this note Section 2.0 of the Tech Specs which covers exactly this topic. I have highlighted the relevant l subsections on the Forced Convection Mode, since we are dealing with issues of primary coolant flow. Here are the definitions from Section 1.0 of our Tech Specs. Safety Limit (SL) - Limits upon important process variables which are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity (10CFR50.36). Limiting Safety System Setting (LSSS) - Settings for automatic i protective devices related to those variables having significant j safety functions, and chosen so that automatic protective action will correct an abnormal situation before a safety limit is exceeded (10CFR50.38). O Lj Phone: "141 '6J-6211 Fw 1417617863 www nmich edu/~mmnn/

~ Q While accurate, these definitions do not capture the practical meaning of the two terms. The term " Safety Limit" has two distinct usages: a) The Safety Limits are calculated quantities based on models and calculations described in the Safety Analysis Report (SAR), which provide bounds within which safe and proper operation of the reactor is assured. In j general, the calculation is conservative in the sense that all of the limits need to be exceeded in order for the design basis condition to occur. For the . Forced Convection System, the design basis is "that the calculated maximum cladding temperature in the bottom of the hot channel of the most compact FNR core (25 elements) will not reach the boiling point of the water coolant". This will not happen unless all four of the variables are violated. We know that the following statement is conservative: Even if the FNR were to operate at all four safety limits - 4.68 MW, 900 gpm,.18 feet and 116 F - the maximum fuel clad temperature would nowhere exceed 235 F, the saturation temperature of water at that depth. For the clad temperature to exceed 235 F anywhere in the core, one or more of the four parameters must exceed its safety limit. The calculation is quite insensitive to the pool depth parameter, so at least one of the other three i -Q must be exceeded. Obviously, if. the inlet temperature exceeds 235 F, then so will the clad temperature, for any flow rate and for zero power. b) The Safety Limits are legal lirnits, which to quote 10CFR50.36: "If any safety limit is exceeded, the reactor must be shut down. The licensee shall notify the Commission, review the matter, and record the results of the review, including the basis for corrective action taken to preclude recurrence. Operation will not be resumed until authorized by the Commission." Clearly, violation of any one of the Safety Limits is a very serious matter. i While the two usages of the term Safety Limit have the same numerical values, to avoid confusion it is important to identify the usage. Once' the Safety Limits have been determined, as described in a), one now needs to determine, for the instruments which measure the variables, set points which will initiate an automatic action if exceeded. The set points are called Limiting Safety System Settings (LSSS) and they are deeigned to prevent Safety Limits from being exceeded. The automatic action is Q usually a scram or a rundown.

1 OPERATING LICENSE AND TECHNICAL SPECIFICATIONS { Ford Nuclecr Racctor j Dtekst 50-2, Licence R-28 Amendment 43:080995 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS' ) 6, LJ i 2.1 Safety Limits 2.1.1 Safety Limits in the Forced Convection Mode Applicability: This specification applies to the interrelated variables associated with core thermal and hydraulic performance in the steady state with forced convection flow. These variables are: Reactor Thermal Power, P Reactor Coolant Flow Through the Core, a Reactor Coolant Inlet Temperature, Ti Height of Water Above the Top of the Core, H Objective: To assure that the integrity of the fuel clad is maintained. Specification: C(g'. 1. The true value of reactor power (P) shall not exceed 4.68 Mw and the true value of flow (m) shall not be less than 900 gym. 2. The true value of reactor coolant inlet temperature (TI) at 2 Mw shall not exceed 116

F.

3. The true value of water height above the core (H) shall not be less than 18 feet while the reactor is operating. Bases: The basis for forced convection safety limits ) is that the calculated maximum cladding temperature in the bottom of the hot channel of 1 the most compact FNR core (25 elements) will not reach the boiling point of the water coolant. ' /~T V ~%

JPLGPMi~TC'LGWLCAELCrl#IPJ) MAL SPECIFICATIONS Ford Nuclect Razctor D:ckat 50-2, Licsnes R-28 Arandment 43:080995 t 2.1.2 Safety Limits in the Natural Convection Mode l Applicability: This specification applies to the interrelated variables associated with core thermal and hydraulic performance in the natural convection mode of operetion. These variables are: Reactor Thermal Power, P Reactor Coolant Inlet Temperature, Ti Height of Water Above the Top of the Core, H Objective: l To assure that the integrity of the fuel clad is maintair.ed. Specification: i 1. The true value of the reactor thermal power (P) shall not exceed 380 kw. 2. The true value of the reactor coolant inlet temperature (Ti) shall not exceed 131 'F. () 3. The height of pool water above the core (H) shall not be less than 18 feet. Bases: l The basis for natural convection safety limits is that the calculated maximum cladding temperature in the hot channel of the most compact FNR core (25 elements) will not reach the boiling point of the water coolant at a l depth of 18 feet. 2.2 Limiting Safety System Settings (LSSS) 2.2.1 Limiting Safety System Setting in the Forced Convection Mode Applicability: This specification applies to the set points for the safety channels monitoring reactor thermal power (P), primary coolant flow (m), height of water above the top of the core (H). and core exit temperature (T.). Pare 7

JPL6T.FIl(CLWLX AlPJ@ YECHNICAL SPECIFICATIONS ~ i Ford Nuclear Rscctor Dockst 50-2 Licenso R-28 Arandrant 43:080995 Objective: \\- To assure that automatic protective action is initiated to prevent a safety limit from being exceeded. Specification: 1. The limiting safety system settings for reactor thermal power (P), primary coolant flow through the core (m), height of water above the top of the core (H), and reactor l coolant exit temperature (T.) shall be as follows: Variable LSSS l P (Max) 2.60 Mw l39 i m (Min) 900 gpa H (Min) 19 ft l T.(Max) 129 *F l Bases: The limiting safety system settings for i forced convection assure that automatic I Protective action will correct the most severe abnormal situation anticipated before a safety limit is exceeded. l 2.2.2 Limiting Safety System Settings in the Natural Convection Flow Mode Applicability: l l These specifications apply to the setpoint for the safety channels monitoring reactor thermal power (P), pool water level (H), and pool water temperature (T). Objective: To assure that automatic protective action is initiated to prevent a safety limit from being exceeded. O i Page 8

r j

PERATING LICENSE AND TECHNICAL SPECIFICATIONS Ford Nuclecr Ragetor Doexat 50-2, Liesnee R-28 i

Amendment 43:080995 i p-Specifications: 1. The limiting safety system settings for reactor thermal power (P), height of water above the to'g of the core (H), and pool water temperature (T) shall be as follows: l ariables LSSS l P (Max) 100 kw l H (Min) 19 ft T (Max) 129 *F Bases: The limiting safety system settings for natural l convection assure that automatic protective action will correct the most severe abnormal si nation anticipated before a safety limit is l exceeded. l I 1 i \\ j I O

NUREG-1138 O Safety Evaluation Report related to the renewal of the operating license for the training and research reactor at the University of Michigan Docket No. 50-2 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation July 191E f-l 's ABSTRACT This Safety Evaluation Report for the application filed by the University of Michigan (UM) for renewal of the Ford Nuclear Reactor (FNR) operating license number R-28 to continue to operate its research reactor has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Comis-sion. The facility is located on the North Campus of the University of Michigan in Ann Arbor, Michigan. The staff concludes that the reactor can continue to be operated by the University of Michigan without endangering the health and safety of the public. O u#a s v

AV 5 REACTOR COOLING SYSTEMS The reactor cooling function is ::erformed by three systems: the p.-imary cooling system, the seconoary cooling system, and the purification system. The controls j and instrumentation associated with these systems are discussed in Section 7. 5.1 Primary Cooling System The components of the primary cooling system are shown in Figure 5.1. These com-ponents include the header and hopper mechanism, the holdup tank, a 1000 gal / min pump, the heat exchanger, and associated piping and instrumentation. The primary coolant system removes up to 2 MW of heat from the core by forced circul6i. ion and maintains the bulk pool temperature at less than 116'F. Tha flow rate is between 900 and 1000 gal / min. A funnel-shaped aluminum hopper, bolted to the bottom of the grid ptate in the core, is designed to direct primary coolant flow from the fuel elements into a header (see Figure 5.2). Its purpose is to reverse flow for convection cooling requirements. The header is held tightly against the bottom of the hopper by a header latching mechanism. A rotating flange allows the header, when released, to swing down and away from the hopper, providing natural convective cooling up through the grid plate and core. l\\/' The header either can be releasea automatically by a low coolant flow rate sig-nal or can be released manually to operate the reactor in a different pool loca-i tion. In the convective cooling mode, reactor operation is limited to power levels below 100 kW. The header latching mechanise and associated instrumenta-tion are described in Section 7. i During forced cooling operations, primary coolant flows down througn the core fuel elements, grid plate, and hooper and then into the header and the holdup tank. The water is pumped by the 1000 gal / min pump through the heat exchanger for heat removal by the secondary coolant system. Primary coolant is returned to the pool from the heat exchanger. The holdup tank is a 1200 gal tank located in the basement belf.w the core. A baffle inside the holdup tac:: provides a flow time lag of ~1 min. The time lag allows the 7.1-s half-life 18N, formed in the core, to decay to a lower level, j minimizing exposure of personnel working near primary coolant piping. b Primary coolant flows from the pump into the heat exchanger shell; secondary coolant flows into an entrance header, through the U-shaped tubes, to the exit header, and then to the cooling tower. Pressure ia the secondary coolant system is greater than in the primary system so any leakage in the tubing will be into the primary coolant, preventing the release of any contaminated primary coolant. 5.2 Secondary Coolino System . f'] The secondary cooling water absoros heat from the primary water in the heat exchanger. The heated secondary coolant water is pumped to a cooling tower UM FNR SER 5-1

h O -- h O, rW U \\Amaryovese/ M I My p 4 v fer* 46 d 4t h !-$-A .s' l im.mup I4 ]"ar. ta d - <gr g A M <ga 4., r . was,m l n W sf Ib 9 'W d,, v @Z~ 27[ w < *w. V 2D W '~' .a.rrd To r.l im,n s e v: we%.,) .. m._,. - w + de i .W Fia une ~ d frase Cef eier (c u.x, x .i. Figure 5.1 FNR primary coolant system UM FNR SER 5-2

~ O 4t'w SW g%o eF mT G N E I TG AN R O TA RFx OL O __~ L F t _C J L ~~ w: O E O B P m U s _g i T n a ) h E D. c M e O A m ( r R 4 e F / p 3 T p N O RR o f, D P h I EO O d /'C I n S O AP a N EU r E HS e P da S e U H S 2 i O p 5 0 er 3 y/ ug i M F S I O/A N G N X E A I T T H IR C A F T O E AL I LM MP / x DX S EI O R RRH T f EEC ) DPA N APT O EOT0 I HHA O 5I

J locatea on the roof of tne reactor cuilding. This is a three-bay, forced-draft cooling tower usea to oissipate the neat to the atmosphere. Eacn of the three cooling tower fans can ce reversea for winter ice removal. Secondary makeup ,\\ water is taken directly from the city water system. 5.3 Coolant Purification System About 25 gal / min of primary coolant is tapped off the primary coolant return line downstream of the flow orifice and flows to the purification system. Cool-ant is pumped past a conductivity cell througn a filter and into the top of either #1 or #2 hot demineralizer units, wnich are located in the demineralizer pit in the reactor basement. Coolant leaves the bottom of the demineralizer unit, passes a second conductivity cell and another filter, and is pumped back to the reactor pool. Except for periods not to exceed 1 week, the pool water is maintained at a resistivity greater than 2 x 105 ohm-cm. The pool water pH is maintained between 4.5 and 7.5. Primary coolant conductivity is recorded continually in the control room. Pool makeup water is supplied by the city water system through a separate filter, cold demineralizer, and conductivity cell and into the primary coolant return line.

5. 4 Conclusion i

The staff concludes that the reactor cooling systems are adequate to prevent fuel element overheating under all normal and likely off normal operating condi-tions and that the coolant purification system can prevent both corrosion and radioactivity problems associated with coolant contamination. l O UM FNR SER 5-4

r. O 7.1.3 Supplementary Control Systems - Primary Coolar.t System The supplementary control systems, also designated as process control systems, provide for operation and regulation of the primary and secondary coolant sys-tems. Included in the process control system are circuits and devices that monitor coolant parameters including temperature, pool water level, and conduc-tivity. Interlocks between the process instrumentation system and the scram system provide positive control of the reactor under all operating conditions. These control systems ensure proper operation of the process related systems and provide information on the status of these systems at the control console -O- - or the instrument panel. During high power operations when the reactor is located in its beam port posi-tion, the primary coolant system removes the decay heat from the core by forced circulation and maintains the bulk coolant temperature below 116*F. A movable 1 header positioned beneath the core provides for forced circulation cooling when i it is held tightly against the bottom of the hopper (header-up position),. direct-ing the primary flow from the fuel elements into the header and subsequently to a heat exchanger for cooling. The header is latched to the hopper by an elec-tromechanical device consisting of a pressure switch that receives its signal from a flow orifice and associated solenoid and is wired to the primary coolant pump circuit. A more detailed description of the primary coolant system is provided in Section 5.1. If the pump flow drops below 900 gal / min or if a power failure occurs, the elec-tromagnet on the heaoer-hopper is deenergized and the header is released, caus-ing it to drop to the lowered position. When dropped to the lower position, the header closes'a microswitch, producing a " header down"' scram signal. Pri-mary coolant flows from the header into the holdup tank and subsequently to the primary pump. From there, it is pumped to the heat exchanger, where the heat can be removed from the primary coolant by the secondary coolant. An itslation valve located at the entrance orifice of ths heidup tank causes idie reactor to be scrameed if the valve is moved from the open position as a result of a loss of p:imary coolant flow. Additionally, a pressure switch located in the holdup i tank also scrans the reactor if a icw pressure condition exists in the tank, which would indicate a substantial leak in the primary cociant system. UM FNR SER 7-4

O j 7.2.2 Process Instrumentation In addition to the nuclear monitoring system described above, there are temper-ature, flow, and water-measuring channels that are additional safety-related instrumentation required by the FNR Technical Specifications for reactor operations. The temperature system consists of temperature detectors (themohns) and a tem-parature recorder that monitor the temperature of the bulk coolant from two positions: 20 ft and 2 ft from the pool surface, primary coolant at the heat exchanger inlet and outlet, and secondary coolant at the heat exchanger inlet and outlet. The recorder has a high temperature switch that is coupled with L the reactor safety system and provides for an automatic rundown of-the safety control rods if the bulk coolant temperature exceeds 129'F. The primary coolant' flow system is required for reactor power operation in excess of 100 kW. The flow system consists of two primary differential pressure trans-ducers that sense (in parallel) flow across of the primary coolant flow orifice. The pressure signal is transmitted to a square root extractor and translated into a flow signal. A primary low flow scram is actuated if the primary flow drops below 900 gal / min and the reactor is at high power operation. Additionally, alarms provide zero flow signals for the high power /zero-flow and header-up/ l zero-flow reactor scrams. Each of the signals is displayed on the digital flow l readout. l The secondary flow transducer senses flow across the secondary coolant flow orifice. The signal is measured similarly to the primary coolant flow signal and then transmitted to a digital flow readout. The pool level monitoring system consists of two float switches. One actuates a local alarm and the other initiates a building alarm when the pool level drops j' 5 and 12 in. respectively. An automatic rundown of the shin-safety rods that is initiated when the pool level decrcases 12 in. is required to be operable before reactor power operation can begin. The primary coolant conductivity measurement system consists of conductivity cells that measure primary coolant conductivity at three locations, one in the domineralizer inlet line and one in each outlet of the two desineralizers. The conductivity is displayed by the conductivity recorder located in the control room. UM FNR SER .7-9

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ML20217E390

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ML20217E390

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