ML20239A053

Expanded Augmented Sys Review & Test Program (Expanded Asrtp) Evaluation of Nuclear Raw Water Sys

ML20239A053
Person / Time
Site:	Rancho Seco
Issue date:	08/18/1987
From:	Akins M, Croley B, Humenansky D SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:
Shared Package
ML20238F564	List: ML20238F561 ML20239A030 ML20239A034 ML20239A038 ML20239A043 ML20239A046 ML20239A053 ML20239A066 ML20239A075 ML20239A082 ML20239A084 ML20239A089 ML20239A104 ML20239A107
References
NUDOCS 8709170053
Download: ML20239A053 (35)

Text

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I li EXPANDED AUGHENTED SYSTEM REVIEW AND TEST PROGRAM (EXPANDED ASRTP)

EVALUATION l OF THE

' NUCLEAR RAW HATER )

l SYSTEM l I

SUBMITTED BY: DATE: 8*/

M. f. AKINS TEAM LEADER CONCURRENCE: ed  % ~ 314 DATE: f-!)- M

[DAVIDHUMENAN' SKY 7 EXPANDED ASRTP PROGRAM MANAGER CONCURRENCE:

/

DATE: E /I/[7 BOB CROLEY DIRECTOR, NUCLEAR CHNICAL SERVICES l

4 O PDR TABLE OF CONTENIS i l

Paae Number 3

1.0 INTRODUCTION

4 2.0 PURPOSE l

5 3.0 SCOPE l

4.0 OVERALL RESULTS AND CONCLUSIONS 6 i l

7 5.0 SPECIFIC CONCERNS 5.1 Acknowledged (Valid) Concerns 9 l 5.2 Open (Potential) Concerns 10 l 1

11 6.0 ATTACHMENTS 6.1 List of Documents Reviewed 12 6.2 Status of RIs 14 6.3 Detailed Observation - Requests for Information 15 l

O EXPANDED AUGMENTED SYSTEM REVIEW AND TEST PROGRAM EVALUATION OF THE NUCLEAR RAW WATER SYSTEM

1.0 INTRODUCTION

The Rancho Seco Expanded Augmented System Review and Test Program

[ASRTP] evaluation effort involres an assessment of the effectiveness of the System Review and Test Program (SRTP] and an analysis of the adequacy of ongoing programs to ensure that systems will continue to function reoperly after restart. The Expanded ASRTP is a detailed systen by system review of the SRTP as implemented on 33 selected systems and an in-depth review of the engineering, modification, maintenance, operations, surveillance, inservice testing, and quality programs. It also conducts a review, on a sampling basis, of many of the numerous ongoing verification ,

and review programs at Rancho Seco. (

Six multi-disciplined teams composed of knowledgeable and experienced personnel are tasked with performing the Expanded ,

ASRTP. Each multi-disciplined team consists of dedicated personnel l with appropriate backgrounds to evaluate the operations, maintenance, engineering, and design functional areas.

Independence, perspective, and industry standards provided by team members with consultants, architect engineer and vendor backgrounds are joined with the specific plant knowledge of SHUD team members.

Each team performs an evaluation on a selected system using the same fundamental evaluation techniques employed by the NRC in the ASRTP inspection. System Status Reports are Jsed as the primary source of leads for the teams. They are augmented with references to available source and design bases documents as needed. Team synergism and communication is emchasized during the process in order to ennance the evaluation. Each team prepares a report for each completed selected system evaluated. This report is for the  !

Nuclear Raw Water (NRH) System. l I

)

I 2.0 PURPOSE The objectives of the Expanded ASRTP evaluation are to (1) assess the adequacy of activities and systems in support of restart and (2) evaluate the effectiveness of established programs for ensuring j safety during plant operation after restart.

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3.0 SCOPE To accomplish the first objective, the Auxiliary Systems team evaluated the NRW system to determine whether:

1. The system was capable of performing the safety functions required by its design bases.

2. Testing was adequate to demonstrate that the system would perform all of the safety functions required.

3. System maintenance (with emphasis on pumps and valves) was adequate to ensure system operability under postulated accident conditions.

4. Operator and maintenance technician training was adequate to ensure proper operations and maintenance of the system.

5. Human factors relative to the system and the system's supporting procedures were adequate to ensure proper system operations under normal and accident conditions.

To accomplish the second objective, the team reviewed the programs as implemented for the system in the following functional areas:

1. Systems Design and Change Control

2. Maintenance

3. Operations and Training j

4. Surveillance and Inservice Testing {

5. Quality Assurance  ;

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6. Engineering Programs The team reviewed a number of documents in preparation for and during the Expanded ASRTP evaluation. This list of documents is found in Attachment 6.1.

The primary source of leads for the team were the problems identified in the NRW System Status Report. Various source documents such as the USAR and Technical Specifications and available design bases documents were reviewed as needed to augment the information needed by the team.

The evaluation of the NRW system included a review of pertinent portions of support systems that must be functional in order for the NRH system to meet its design objectives.

4.0 OVERALL RESULTS AND CONCLUSIONS The more significant issues identified certaining to the adequacy of .

the SRTP and the effectiveness of prc 1ms to ensure continued safe operations after restart are summari: a below. The summary focuses on the weaknesses identified during tne evaluation. Attachment 6.3 provides detailed findings by providing the Request for Information (RI) forms that are used by the Expanded ASRTP teams to identify potential concerns during the evaluation. Section 5.0 lists the i specific concerns identified by the teams. The numbers in brackets i after each individual summary or concern refer to the corresponding RIs in Attachment 6.3.

l The conclusion reached from the review of the Nuclear Raw Water  !

System is that there is enough inconsistencies in the design calculations and the surveillance procedures to raise a concern i about the ability of the system to meet its intended functional requirements.

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QVERALL RESULTS AND CONCLUSIONS (Continued) 4.1 NUCLEAR RAW WATER (NRW) FLOH RATES 4.1.1 A review of the system P&ID M-544, Rev. 20, revealed a lack l of any flow measuring instrumentation for the system. The Updated Safety Analysis Report (USAR) and the System Design Basis (SDB) describe the minimum flow rates required through the system and the individual system components.

A review of the Quarterly Surveillance and Inservice Test L Procedure (SP.203.07A/B/C/D. Rev. 18) revealed that only the c total system flow rate is estimated. A review of inspection l l

data revealed that the method used to determine flow rate l yielded data that, in a majority of cases, were beyond the region of pump runout. No documented evidence was able to be produced to show that this was ever questioned.

The flow rate in the system is determined by measuring spray header pressure and referring to a flow rate versus pressure graph in the procedure. Nuclear Engineering was not ab'a to provide a source for the curve, nor have calculations been produced to support the curve. A concern exists as to the '

validity of flow rates in the system throughout the life of the plant and more specifically that individual components such as the Emergency Diesel Generators (EDG) Jacket Water Coolers have ever received their minimum required flow.

[RIO01] [RIO33] [RIO56].

4.1.2 A review of all of the NRW calculations failed to show that adequate flow has been provided through any of the lube oil or room coolers other than through the Nuclear Service Water (

Cooler.

No calculations could be produced that verify, when the system was designed, adequate flow could be provided to all system components. Individual loop flow rates have been calculated based on the required capacity without considering supply pressure. Because the flow through these loops provide cooling to safety related pumps, lube oil coolers and pump room coolers, improper sizing of the system will reduce the ability of the system to meet the functional support needs of the safety related components. [RIO32]

(RIO25]

OVERALL RESULTS AND CONCLUSIONS (Continued) 4.1.3 Numerous inconsistencies exist in the calculations which would result in the safety related pump cooling systems being l improperly sized. Similarly, a variety of inlet water temperatures are used in calculations; for example, 87'F for the I Nuclear Service Cooling Water Heat Exchanger, 82*F for the EDG l Jacket Water Cooler. The Updated Safety Analysis Report (USAR) <

indicates that 95*F should be used. This agrees with a hi-temp alarm of 95*F for the spray ponds. [RIO34] [RIO64) 4.1.4 The commitment to balance the NRW system flows prior to startup has been completed. A review of the procedures and observations revealed concerns about the qualifications of the testing personnel, calibration of the equipment and the training for use and care of the instrument. No documents were able to be produced by System Testing personnel or Training to remove this concern.[RIO20]

4.2 NRH Pond Recirculation Subsystem 4.2.1 The System Status Report (SSR) for the NRW System identified several instances where corrosion, algae and sediment have entered the system and caused plugging of the lube oil coolers and room coolers. No documents could be produced which indicate that a root cause analysis was ever performed to determine the mechanism (s) which create the problem identified in the SSR.

[RIO99]

The surveillance procedures developed to monitor cooler plugging do not contain acceptance criteria for determining when pluggir.g is beyond allowable limits. [RIOS6]

4.2.2 A review of the cond recirculation system revealed a lack of engineering documentation to support the design. Since the only cleanup and chemical control is provided via the recirculation system, no documents could be produced that show that :he syst a l

was designed to meet the requirements of the USAR. [RIO271 It was observed that misoperation of the existing system could occur because compenents are mislabeled and the Spray Pond level switch is improperly sized for controlling the system. [RIO55]

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[RIO26] [RIO98]

4.3 Programmatic 4.3.1 Several instances were found where components and systems were released and declared operational prior to closure of the associated Engineering Change Notices (ECN). [RIO21]

4.3.2 Material for safety related components was purchased as commercial grade without the required dedication documentation.

[RIO58]

i 5.0 SPECIFIC CONCERNS A list of the specific concerns the Expanded ASRTP team believes are new concerns not previously identified for resolution follows:

5.1 Acknowledged (Valid) Concerns 5.1.1 The reisting NSRW system does not provide any means for ,

dettn: Ing the flow rates through system components to ensure that minimum flow rate requirements are being met for safety related components. [RI 001]

5.1.2 The NSRW system flow balance and associated data could be erroneous due to lack of calibration, training, and procedures l for the controlotron flow meter. [RI 020]

5.1.3 Emergency pump room, makeup pump lube oil, and High Pressure Injection Pumps A & B lube oil coolers were addcd to the NSRW system without a proper review of the heat load effect on the operating system. [RI 025]

5.1.4 The NSRW recirculation and filtration system may not be providing the necessary filtration to meet USAR commitments since no design bases or calculation exist to support the design nor is there any instrumentation to verify operability. [RI 027]

l 5.1.5 The original design calculations for the NSRW were incorrect by failing to balance all parallel flow paths simultaneously.

[RI 032]

l 5.1.6 The NSRW fiow rate versus pressure graph used in SP 203.7 (NSRW Surveillance and Inservice Test Procedure) to verify system l operability is unverified. [RI 033]

5.1.7 Based on design information, if the NSRW spray pond temperature l

reaches it alarmed temperature condition of 95F, several safety l related components may not be able to perform their intended j cooling function. [RI 034]

5;1.8 Safety related equipment and components are not properly tagged or labeled. [RI 055] 4 5.1.9 Surveillance Procedure SP 203.07 "NSRW and NSCW Systems Quarterly Surveillance and Inservice Inspection Test" was revised without including all acceptance criteria needed to determine system operability. [RI 056]

5.1.10 Several NCRs were not written against NSRW ODRs with nonconforming conditions. [RI 057]

5.1.11 Unmedicated commercial grade lube oil coolers have been recently installed in the safety related NSRW system. [RI 058]

5.1.12 The low-level alarm setpoint for the NSRW spray pond is below  !

the minimum level required to meet the seven day, post LOCA, l criterion. [RI 098]

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5.1.13 The solutions proposed for several NSRW SSR problems do not resolve the problems, nor do they address the fluid dynamics of the system which indicate that the present flow may not be adequate with no plugging or fouling. [RI 099]

5.2 Open (Potential) Concerns None l

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/ I 6.0 ATTACHMENTS l

6.1 List of Documents Reviewed j Status of RIs  !

6.2 6.3 Detailed Observations - Request for Information I i

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LIST OF DOCUMENTS REVIEWID STP.1031 B, 1084 ECN A3795, A3795A,B,C,0,E,F, R0328, A4905, A5619, A5651 LER 81-16, 87-36 00R 86-263,86-386, 86-388,86-366, 86-276,87-282, 86-167,86-396, 86-386,86-394, 86-457,87-220, 87-241,87-404, 87-506,87-560, 87-638,87-746, 87-753,87-760, 87-764 f A0R 50-312-74-7, 50-312-74-9 E-480A&B Vendor Manual & Bid Specs A-529A,B,C,0,E Vendor Manual & Bid Specs P-472A&B Vendor Manual & Bid Specs F 476A&B Vendor Manual & Bid Specs F-475A&B Vendor Manual & Bid Specs P-471A&B Vendor Manual & Bid Specs P-238A&B Vendor Manual P-261A&B Vendor Manual P-236 Vendor Manual P-291 Vendor Manual Controlotron Vendor Manual Controlotron QA Manual Chemical / Radiation Logs 01-01-84 through 12-30-84 P0 RS-37743, RS-93501 Master Equipment List (MEL)

RJR Letters74-440, 74 445 MSRC Heeting Agenda 12-05-74 TS/600/33 Calibration Records for LSL-47001 USAR Technical Specifications Process Standards System Operating Procedures System Design Bases Annunciator Procedures Casualty Procedures Chemistry Manual Station Manual QA Hanual System Status Reports Bechtel Design Calculations Z-NRW-H0236 Bechtel Design Calculations Z-NRW-H0246 Bechtel Design Calculations Z-NRW-H0237 Bechtel Design Calculations Z-NRW-H0238 Bechtel Design Calculations Z-NRW-H0157 Bechtel Design Calculations Z-NRW-H0243 Bechtel Design Calculations Z-NRW-H0245 Bechtel Design Calculations Z-NRW-H1729 Bechtel Design Calculations Z-NRW-M1735 Bechtel Design Calculations Z-HVS-H2258 Bechtel Design Calculations Z-NRW-H2260 ATTACHMENT 6.1 LIST OF DOCUMENTS REVIEHED (Continued)

P& ids M-544, Rev. 20 AP.22, 33, 44, 46, 306 System Training Manuai -

System Lesson Plan SP.203.07A/B/C/D, 206.03A ,

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HR '69886, 76082, 62926, 63190, 76081, 98766, 77882, 69884,.76090, 98769, 69885, 69882, 76094, 78839, 78771, 69883, 98768, 76093, 7784, 60503, 83834, 98770, 76089, 54052. 54047, 64809, 65027, 60501, 49753, 64807, 65028, 60503, 49754, 56338, 64808, 71209, 49752, 54050, 54051, 60016, 63052, 77929, 002340, 014103, 019929, 54135, 66873, 68902, 014115, 014148, 71440, 7)910, 92880, 98730, 98766, 98768, 98770, 99956, 104680 105120, 107435, 107824, 110755, 111359, 111840, 111843, 113174, 113175, {

113176, 113981, 113982, 113983, 114587, 114611, 114619, 114900, 115333, 115334, 115341, 115346, 115719, 115761, 115760, 115795, 115809, 115864, 116232, 116662, 117254, 117255, 117256, 117257, l'i7258, 118036, 118037, 118038, 121770 NCR 1589, 2039, 6264, 4780, 2898, 6664, 3634, 4489, 3834, 4303, S2522 through $2527, S3178, S4565, S4726 S4727, 4913, S4969, S5023, 5242, 5417, S5425, S-5645, S5464, 5499, 5584, S5645, S6103 Engineering Report #ERPT-E0187 GVC-87-056 NUREG 800, Rev. 2 Precursor Review Task Final Report RIDR QA #1236, #3590 QA Audit 0-494, 87-10, 87-21, 87-23, 87-24 M30.01-5 Approved Suppliers List I

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ATTACHMENT 6.1

STATUS OF RIs Attachment 6.2 provides RI status as of this report date. An RI is considered closed if the Team Leader was convinced a potential concern war act valid or not significant enough to be an RI. An RI would also be cic , if requested information was provided. All other RIs are open.

Acknowledged RIs are open RIs that have been accepted as valid by the responsible organization and nave been stated as concerns in Section 5.0.

RI NUMBER STATUS 001 ACKNOWLEDGED 020 ACKN0HLEDGED 021 CLOSED 025 ACKNOWLEDGED 026 ACKNOWLEDGED 027 ACKNOWLEDGED 032 ACKNOWLEDGED 033 ACKN0HLEDGED 034 ACKNOWLEDGED 055 ACKN0HLEDGED 056 ACKNOWLEDGED 057 ACKNOWLEDGED 058 ACKNOWLEDGED 063 CLOSED 064 CLOSED 098 ACKNOWLEDGED 099 ACKNOWLEDGED ATTACHMENT 6.2

DETAILED OBSERVATIONS - REOUEST FOR INFORMATION During an evaluation, all potential concerns are documented on Request for Information sheets (RIs) that are sent to the responsible organization to receive their input concerning the potential concern. RIs are also used to request information that the EASRTP team is having difficulty obtaining. 3 These RIs are consideret arafts throughout the entire evaluation until they become part of the report. Responsible organizations can accept the potential concern as valid or they may disagree with the >

l potential concern. If they disagree, they can submit information that convinces the EASRTP team members that the potential concern is not valid, or they may redirect the EASRTP members to better focus the concern. RIs developed during the system evaluation comprise this section of the report.

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1 ATTACHMENT 6.3 l

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REQUEST FOR INFORMATION (RI)

RI NO:__QQ1 SYSTEM CODE: NSRH ISSUE DATE: 7/17/87

SUBJECT:

ADEOUATE FLOW TO D/G JACKET HATER COOLERS DEPARTHENT: NED _ COORDINATOR: R. LAHRENCE TEAM LE8, DER: 4.J. AKINS l POTENTIAL CONCERN /0VESTION:

The existing NSRH Configuration does not provide an accurate means for  ;

determining that the Diesel Generator jacket water coolers (0/G J.H.Hx.) are provided with a minimum of 1000 gpm as required by design (ref. .

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Thermaxchanger specification sheet, see calc. Z-NRH-H0236. file N19.01) l

- Flow diagram H-544 rev. 20 does not identify any flow measuring instruments anywhere in the NSRH system. It does provide a pressure indicator (local) on the pump discharge line and a remote pressure indicator (Control Room) on the spray discnarge. These do not provide measurement of flow through aither the system or the D/G J.H.Hx.

. NSRH Quartsrly Surveillance and Inservice Inspection Test Procedure (SP 203.07 A/B/C/D/, Rev. 18. dated 7-25-86) does no' identify any measurement step for D/G J.H. Hx. flow.

. This procedure does provide a method for estimating total system flow. The flow grapn turve provided for this purpose has no 4 identifiable foundaticr. and thus the resultant flows are questionable.

. Attachment A shows a plot of the system flow versus pressure from the data contained in the surveillance records for the system. As can be seen, most of the data points are bevond pump l runout (based on mfgr. pump curve) which is physically impossible.

. The pumo/ motor vendors performance curves using pressures and motor ampere readings indicate system flows other than those estimated.

• A walkdown of the system confirmed that no flow measuring devices exist anywhere in the system. j t

. The D/G monthly surveillance Test Procedure SP 206.03A, Rev. 16, dated 03-07-86 does not address HSRH flow.

ATTACHMENT 6.3

i POTENTIAL CONCERN /0UESTION (Continued)

. A visual inspection of the D/G local control panel revealed that a flow indicator for measuring jacket water flow exists on the panel. However, it is not connected. No one could remember why it was not connected or when it was disconnected.

- Discussions with some operators and I&C indicate that the flow indicator has never been connected. The operators felt that it should be.

In summary, there is no way for any operator to determine flow rates through '

i D/G J.H.Hx. at anytime. There appear to be no procedures in place to monitor or check flow. The existing system flow calculating procedure is crude and yields inaccurate data.

One potential impact stemming from this is:

During a loss of offsite power, concurrent with failure of 1 D/G, the potential for losing the second D/G on high Jacket Water temperature caused by insufficient NSRH to the D/G J.H.Hx. exists.

l ATTACHMENT 6.3 4

I REQUEST FOR INFORMATION (RI)

RI NO: ..020 SYSTEM CODE: NSRW ISSUE DATE: 7/22/87

SUBJECT:

FLOH READINGS ON NSRW TAKEN WITH THE CONTROLOTRON FLOWMETER DEPARTMENT: SRTP COORDINATOR: J. ITTNER TEAM LEADER: H.J. AXINS POTENTIAL CONCERN /0UESTION:

1 A concern exits that the NSRW flow balancing and flow rate data will be of acceptable accuracy. Procedures STP-1031A NSRH Loop A and B flow balance require flow to be measured. The flow readings for these are being obtained using a Controlotron Ultrasonic Flowmeter. Based upon a review of procedures, calibration data, QA documents and training records, it was found that the plant personnel could not provide:

. procedures for the care and use of the meter a documents showing approval of the Vendor manual

. records of any training of testing personnel on the use, operation and care of the device. l Field observation has identified that the tracks for one of the meter point.

were installed improperly and would yield erroneous flow readings. Also, discussions with personnel installing the flowmeter indicated their uncertainty on bow to connect and reed the instrument.

Because the readings are being used to verify the flow balance of the NSRW system, wnicn is a Safety-Relatea System, and because there are no flow devices in the entire NSRH system, it is important that the flow data <

collected and verified is accurate for baselining the system. l 1

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ATTACHMENT 6.3

REQUEST FOR INFORMATION (RI)

RI NO: 021 SYSTEM CODE: NSRW ISSUE DATE: 7/23/87

SUBJECT:

POSSIBLE VIOLATION OF PROCEDURE AP.44 DEPARTHENT: OPERATIONS COORDINATOR: R. MACIAS TEAM LEADER: M.J. AKINS j POTENTIAL CONCERN /00ESTION:

System was turned over to operations while open ECNs exist. This could be in violation of procedure AP.44 (Plant Modifications - ECN Implementation, Rev.

11, dated 03-17-87) unless an interim release has been documented. A review of procedure AP.44 states that a system cannot be declared operable until all open ECNs for the system have been closed or an interim release.

• SSR problem #8 states that Pressure Indicaters will be installed across the High Pressure Injection, Makeup and Decay Heat Removal pump lube oil coolers to monitor plugging trends.

• ECN A3795 was generated to respond to SSR #8. It is still ooen in SDC yet the system is in operation and plant personnel were unable to produce documents that confirm that the requirements of AP44 have been satisfied.

• This concern is closed because the problem will be handled through an Engineering Action Request (EAR).

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l ATTACHMENT 6.3

REQUEST FOR INFORMATION (RI)

RI NO: 025 SYSTEM CODE: NSRH ISSUE DATE: 7/23/87

SUBJECT:

LUBE OIL COOLERS AND PUMP ROOM COOLING HEAT LOADS DEPARTHENT: NED COORDINATOR: RON LAHRENCE TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /0UESTION:

Modifications were made to the NSRH System without proper review of effect to the operating system. Because of this the system may be unable to meet its functional objective of supplying cooling to safety related equipment.

. Based on a review of the NSRH P & ID M-544 and a review of original design calc. no. Z-NRW-M0246, it has been found that the heat load generated by the emergency pump room cooler, make up pump, High Pressure Injection A & B were not taken into consideration when designing the NSRH system. It appears that the system was modified without properly analyzing the effect ot' the change.

Since NSRH is the ultimate heat sink, the heat loads generated by the aforementioned should have been evaluated prior to implementation of an" ]

system changes.

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l ATTACHMENT 6.3 l

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REQUEST FOR INFORMATION (RI)

RI NO: 026 SYSTEM CODE: NSRW ISSUE DATE: 7/22/87

SUBJECT:

LQW LEVEL SWITCH-SDRAY POND A (LSL 47001)

DEPARTMENT: SRTP COORDINATOR: J. ITTNER TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /0UESTION:

NSRW Circulating Pump P-476A is not properly controlled by LSL-47001. P-476A i

may run without water or may not run sufficiently to keep the pond clean.

The NSRH Circulating Pump P-476A is not controlled adequately. Review of the NSRH SSR, prccess standards and calibration records indicate:

. Problems #12 and #1S in the NSRH SSR still exist.

. Present level switch (LSL-47001) which controls the circulating pump has a range of 0-60" water.

. Process standards AP.107 list the setpoint of LSL-47001. or 0-67" H O 2 +

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. ECN A-S619 replaces LSL-47001 and is not planned to be implemented.

before restart.

Since the recirculation system is the only means provided for cleaning the spray pond, it is important that it is functioning properly to ensure that debris, sediment, slime, etc. are removal from the system to preclude plugging of components. The system should be capable of the designed function prior to restart.

This concern was closed since it is already covered adequately in the SSR program.

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ATTACHMENT 6.3

i REQUEST FOR INFORMATION (RI)

RI NO: 027 SYSTEM CODE: NSRH ISSUE DATE: 7/23/87

SUBJECT:

NUCLEAR RAW WATER SPRAY POND RECIRCULATION AND FILTRATION SYSTEM NED COORDINATOR: RON LAWRENCE DEPARTMENT:

TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /0UESTION:

There is no design bases or calculations to support the present installed recirculation and filtration system equipment. Further, the system that is installed may not be providing the necessary flow rate for proper filtration {

to meet USAR commitments, nor is there any instrumentation to determine equipment operability.

. USAR Volume IV Section 9.4.2.1 states that each of the ponds has a separate chemical feeding and filtering system for continuous cleaning of the water.

. Bechtel Design Calculation Z-NRW-M0157 originally called for a skimmer system that was capable of recirculating a volume equal to the pond width times the pond length times a depth of one inch in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time span. Using a 1-1/2 inch pipe and taking suction at one corner of the pond and discharging to the opposite corner, established the need for a pump with a rated discharge head of 30 feet. The volume requirements set the pump capacity rating at 20 gna. The calculation does not I address the effects of filtration. One year after the calculation was )

complete the pump rating and pipe size were changed for no apparent justification. The original project engineer was interviewed and stated that ne could not remember :he reason for the change. The new figures  !

were for six inch piping and a pump rated at 500 gpm at 30 feet of head. !

. Telephone conversations held with the filter supplier indicates that SMUD ordered a 500 gpm filter and did not request a filtration system sized for the ponds. ,

. At the present rating of 500 gpm recirculation flow would completely recirculate the pond in approximately four days.

. Telephone conversations held with Sacramento County Water Quality Board Design Engineers indicate that for proper filtration of any body of water, the entire volume must be filtered in each 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.

ATTACHMENT 6.3 I

POTENTIAL CONCERN /0UESTION (Continued):

. Telephone conversations held with the filter manufacturer confirm this 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> recirculation figure. l

= A visual inspection of the present recirculation / filtration system revealed that there are no pressure, flow, or temperature instruments on

'the system to confirm satisfactory operation.

- Discussions with operators indicate that their only method for l determining flow was to listen to the filter for sounds of water flow. l i

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ATTACHMENT 6.3

REQUEST FOR INFORMATION (RI)

RI NO: 032 SYSTEM CODE: NSRH ISSUE DATE: 7/24/87

SUBJECT:

INCORRECT DESIGN FLOWS FOR COMPONENTS IN PARALLEL FLOH PATHS DEPARTMENT: NED COORDINATOR: RON LAHRENCE TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /00ESTION:

The original design calculations Z-NRW-M0246 and others were incorrect by failing to balance all parallel flow paths simultaneously. The design l calculations traditionally only verify the main flow and pressure drop to I satisfy the pump requirement, and assume that as long as the main flow is adequate, the side stream flow requirement will be met. In this case, the inadequate flows through the side stream coolers could result in the damage of the safety related equipment, and thereby challenge the integrity of the safety feature of the plant.

. Nuclear service cooling water heat exchanger (1)

Flow - 15,000 gpm 164.62 x E6 BTU /hr Source - Z-NRW-M0246  ;

• Diesel generator cooling water heat exchanger (1)

Flow - 1000 gpm 9.95 x E6 BTU /hr Source - Z-NRH-M0246

• High pressure injection cump and makeup oump lube oil cooler (2)

Flow - IB gpm 14.11 x E6 BTU /hr i Source - Z-NRH-M0246, B&W 1etter B/10/70

- Decay neat removal pump bear;."a oil cooler (1)

F1'ow - 10 gpm 1000 BTU /hr i Source - X-NRW-M0246, B&W letter 6/3/70 )

1

. Emergency pump room air coolers (3)

Flow - 35 gpm 170,000 BTU /hr Source - Vendor Data Sheet, Marlo Coil works, 4.12.71.  ;

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. Reactor building spray pump bearing oil cooler (1) I l

Flow - 10 gpm (assumed) l Source - Z-NRH-M0246 l l

ATTACHHENT 6.3 I

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)

POTENTIAL CONCERN /00ESTION (Continued):

All flows through the components and coolers are arranged in parallel flow path. The only design calculation (A-NRW-M0246) available prior to the restart program is the main flow through the nuclear service cooling water heat exchanger, where the pressure drop and flow were calculated to meet the nuclear service raw waier ptmp head requirement. The calculation also include pressure drop through the diesel gerarator cooling water heat exchanger. All parallel flow paths should be balanced simultaneously to determine flow rate through each path for pump head requirement.

. Survey of all NRW design calculations fails to show any calculation to demonstrate adequate flow through the components and coolers other than the nuclear service cooling water heat exchanger.

. Recent calculation .Z-NRW-M2260 balancing nuclear service cooling water heat exchanger, decay heat removal pump gearing oil cooler, and emergency pump room air cooler in a parallel flow paths, shows flows less than the design value of 10 gpm for the decay heat removal pun.p bearing oil cooler, and 35 gpm for the emergency pump room air cooler.

. Flow tests dated June 24, 1987, indicated all parallel side stream flows ,

through the coolers were below the design flows as shown above, except  !

l the diesel generator cooling water heat exchanger.

. Flow tests also showed that there were inadequate flows through the side streams even with the valve at the main flow through the nuclear service cooling water heat exchanger completely closed.

In summary, the original design calc'ulations Z-NRW-H0246 and others were incorrect in failing to balance all parallel flow paths simultaneously.

Instead, only the main flow of 15,000 gpm through the nuclear service cooling water heat exchanger was verified to meet the pressure drop requirement of the cump. It appears the calculation assumes that there will always be  :

sufficient flow through the side stream wnicn is approximately 100 gpm oy j throttling the valve at its main flow. The assumption is true only when the l side stream flow resistances are relatively low. I i

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ATTACHMENT 6.3

REQUEST FOR INFORMATION (RI)

RI NO: .033 SYSTEM CODE: NSRH ISSUE DATE: 7/23/87

SUBJECT:

SP.203.07 A/B/C/D SRTP COORDINATOR: JOHN ITTNER l DEPARTHENT:

TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /0VESTION:

Unverified data is being used to determine system flow rates in a safety related system. On October 6, 1974 the "A" side nuclear service cooling water surveillance, SP.203.07C, was performed, and the results turned over to the plant mechanical engineer for evaluation. Seventeen days later the system engineer determined that the cooling water pump had not met the required flow rate. On November 5, 1974 an abnormal occurrence report was filed with the Atomic Energy Commission. In the report SMUD decided that the

surveillance procedure would be revised to include " acceptance criteria" so that the operator who records the data would be able to evaluate the results. On November 18, 1974, Mr. Rodriguez directed the Plant Hechanical Engineer, to handle revising the procedure. On December 4, 1974 the procedure was revised to include Enclosure 6.7 (Attachment #2), a flow versus spray header pressure graph. The plant personnel were not able to provide verified data supporting the generation of the graph. This graph is still being used to determine total flow.

Since this graph is the only means for determining system flow it is important that it be as accurate as possible and represent all possible system conditions which may exist at the time of surveillance.

ATTACHMENT 6.3

) i REQUEST FOR INFORMATION (RI) ,

4 RI NO: 034 SYSTEM CODE: NRW ISSUE DATE: 7/23/87

SUBJECT:

NUCLEAR SERVICE RAHEWATER TEMPERATURE INCONSISTENT WITH TEMPERATURES SPECIFIED FOR GOMPONENT MANUFACTURES i NED ~ COORDINATOR: R. LAWRENCE DEPARTMENT: 1 TEAM LEADER: M.J. AKINS f POTENTIAL CONCERN /0UESTION Based on the design information, if the spray pond' temperature reaches its l

l alarmed temperature condition of 95F, several safety related components may not be able to perform their intended cooling function.

USAR 9.4-20 states ".. 95F was the upper temperature limit for cooling water temperature specified to componant manufacturers..." <

System design basis document, page 11 of 24 indicates 95F as the maximum spray pond outlet design temperature.

Inlet temperature of 87F was specified for the Nuclear Service Cooling Water Heat Exchange.

Inlet temperature of 85/ @ 12 GPM, and 95F 6 GPM for the high pressure injection and rake up water pumps.

If the spray pond-temperature ever reaches it alarmed condition of 95F, the heat transfer capability of the above identified components may not be sufficient to remove the required amount of heat from the safety related components they cool.

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ATTACHMENT 6.3 i

5 '

4

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REQUEST FOR INFORMATION (RI)

RI NO:. 055 SYSTEM CODE: NSRH ISSuF DATE: 7/27/87

SUBJECT:

PHYSICAL 10ENTIFICATION OF HOTORS. PUMPS. AND INSTRUMENT TION DEPARTMENT: OPERATION COORDINATOR: R. MACIAS TEAM LEAN R: M . J . AK_LN3 l

POTENTIAL CONCERN /0UESTION:

Equipment and components necessary for system operations are not properly tagged or labeled. Since field mounted equipment (motors, pumps, instruments, etc.) are not individually identified, the validity of identification for operation or repair of said equipment on approved documents (open and those closed out) are in question. The misoperation, replacement, or repair of designated equipment for any field action is then in question. The mis-operation, exchange in shop repair, and calibration / modulation of specific equipment could affect the plant functionality and create inaccuracies in the Master Equipment List.

. The existing NSRH system motors, pumps, and instrumentation, etc., lack a physical identification of equipment. In a walkdcwn, two pumps were found having been identified by hand painting the designation.

. Of the two hand painted pump designations, one is found badly faded, a the second was found i m adable due to the peeling of the base paint.

. The NSRH Circulating M tr Pumps P476A and P;476B have the same identification: Westinghouse, 7.5 hp, model number 71015449, SN 7105.

Both plates rsad the same.

• In a supplemental walkdown, equipment was randomly selected without regard to system or physical location, and no designation on the small equipment was found. An exception was found, in that the very large pieces of equipment were found to ha7e been stencilled with a designation.

The following tagging and labeling problems were found in the CR/TSC Heating  :

and Ventilating Syste.n.

. Manual Isolation Valves HVS-007, HVS-009 -Oll,-017, and -019 (P&ID M504 F-4) lack valve ID tags.

! . Breaker 2A323 label indicates the load on that breaker but does not l indicate that it is breaker "2A323." Breaker 10 is written on the door with " Magic Marker."

. Breaker 1A324 label does not list HV-54727 as one of its loads as indicated on drawing E-107 Sheet 27.

ATTACHMENT 6.3

i l l i

l' REQUEST FOR INFORMATION (RI) i RI NO: ..056 SYSTEM CODE: NSRW ISSUE DATE: 7/28/87 l

SUBJECT:

MCRS 4969 DISPOSITION AND CORRECTIVE ACTION l

DEPARTHENT: _SRTP COORDINATOR: JOHN ITTNER TEAM LEADER: M.J. AKINS POIENIIAL_ CONCERN /0UESTION: 1 SP 203.07 has not been revised as required by the disposition to NCR S-4969.

The disposition of NCR 4969 requires revision of SP.203.07 to include the recommendations identified in Z-NRH-M1735. A review of the SP and the calculation indicate that the SP has been changed to require measurement of the pressure drop across the NSCH Heat Exchanger tubes and has included the acceptance criteria for the NSCH Heat Exchanger tube delta P and the NSRH spray nozzle delta P. However, the calculation also addresses that the cumulative pressure drop across the heat exchanger tubes and spray nozzles should not exceed 5 psig. This cumulative pressure drop has not been addressed in SP.203.07. Case I in the conclusions section of the calculation provides an exarple of how individually each component (tubes and nozzles) may be within the acceptance criteria, but the cumulative pressure drop exceeds the acceptance criteria and, therefore, per the calculation the

" system should be shut down and investigated." This case would not have been identified using the Existing SP.

1 ATTACHMENT 6.3 1

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REQUEST FCR INFORMATION-(RI) i RI NO: 057 SYSTEM CODE: NSRH ISSUE DATE: 7/28/87-

SUBJECT:

NONCONFORMANCES IDENTIFIE0 ON 00Rs DEPARTMENT: LICENSING COORDINATOR: JERRY DELEZINSKI TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /0UESTION:

NCR's were not written against 00Rs with nonconforming conditions thus violating QAPl7. Of the 20 NRH 00Rs issued during 1986 and 1987, eight '

required NCRs to be written per the requirements of QAPl7. However, NCRs .'

were identified for only four of those eight ODRs'. The other four (86-263,86-338, 86-366, and 87-753) do not appear to have had NCRs written against them even though nonconforming conditions existed.

A review of the closure process for these four 00Rs indicates that some control exists to assure that the necessary corrective action is completed.

However, the level of review for both the discrepancy description and disposition do not appear as extensive as is required for an NCR. In fact, it could not be determined who reviews 00Rs to assure that the response i adequately addressed the original discrepancy.

The existing procedure, AP.22, Rev.12, requires the 00R originator to determine if an NCR should be initiated. Apparently AP.22 is being revised to require Licensing review of discrepancies for NCR considerations but this revision has not yet been issued.

During this 00R review it was also identified that the 00Rs were not being stored properly as QA records in violation of RSAP-0601. The 00Rs were not filed in fire-proof cabinets prior to turnover to RIC.

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I ATTACHMENT 6.3 l 4

REQUEST FOR INFORMATION (RI)

RI N0:_ 058 SYSTEM CODE: NSRW ISSUE DATE: 7/28/87

SUBJECT:

PROCUREMENT OF LUBE OIL COOLERS FOR ECN A-3795 DEPARTHENT: NED COORDINATOR: RON LAWRENCE 4

TEAM LEADER: H.J. AKINS POTENTIAL CONCERN /OUESTION:

Unqualified materials which may not be able to meet their intended functional requirements are being used in safety related systems. Per RIDR, OA #1236, the lube oil coolers were purchased as Commercial Grade parts used in OA Class 1 safety applications is required by 10CFR21. Standard industry practice defines dedication as the identification of attributes critical for the component to perform its safety function and the determination /

implementation of methods to assure that these attributes are verified as acceptable, i.e., visual inspection, testing.

Plant personnel have not been able to produce documentation to substantiate dedication of these coolers. The DBR briefly indicates that the difference in heat removal between the old and new coolers would not affect design. A letter from the cooler supplier states that the coolers are dimensionally interchangeable. However, this documentation does not fully address the cooler dedication.

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I ATTACHMENT 6.3 l

REQUEST FOR .(NFORMATION (RI)

RI NO:. 063 SYSTEM CODE: ESRW ISSUE DATE: 7/28/87 SUBJ ECT'. DESIGN GUIDE REFERENCE TO NO EXISTENT DESIGN CRITERIA DEPARTMENT: NED COORDINATOR: RON LAWRENCE l

TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /00ESTION:

I I

The procedures used for determining Environmental Qualifications via the Design Guides refers to another criteria document that does not exist.

• Design Guide 5204.51 (approved 09-23-85) " Limitations on use of electrical equipment and materials in a Nuclear Containment," paragraph 5.3. and 6.1, Reference Design Criteria 5101.4 " General Design Criteria, Environmental," as the guide fc'r material qualification. Design Criteria 5101.4 is said by Nuclear Engineering, to not exist.

= Engineering Report #ERPT-E0187 " User Manual for the Rancho Seco Environmental Equipment Qualification Program," approved in Engineering dated March 27, 1986, was offered as the alternate to the Design Criteria 5101.4 by the Environitental Qualification Engineer in the Electrical Engineering group.

• This concern was closed because resolution will be handled through an Engineering Action Request.

ATTACHMENT 6.3 1

RE0 VEST FOR INFORMATION (RI)

RI NO:__064 SYSTEM CODE: NSRW ISSUE DATE: 7/28/87 l

SUBJECT:

PUMP R00H COOLER ANALYSIS AT REDUCED NRW FLOW _

DEPARTMENT: NED COORDINATOR: RON LAWRENCE TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /00ESTION:

Verified and approved calculations are being used to support restart which contain fundamental errors. 2-HVS-M2258 was performed to justify 16 gpm instead of 35 gpm and an inlet temperature of 95*F for nuclear service raw water.

l

. It appears that there is no heat balance between the water side and the  !

air side of the cooler, as evidenced by the results of the calculation page 12 and page 9.

. The water side temperature increases from 95'F inlet to 134*F outlet, to give a total heat removal capacity of the cooler 312,000 BTU /hr (page 9), which contradicts water side heat load of 312,000 BTU /hr.

. It should be recognized that 176,273 BTU /hr is a maximum design heat load, not the heat removal capacity of an existing cooler with a given surface, nor should it be assumed so.

. A room air temperature of 130*F under these conditions is not acceptable, as it exceeds the normal harsh environmental temperature of 108'F (Engineering Report ERPT-E-187).

This was closed because the calculation was redone to remove the calculation errors. l l

ATTACHMENT 6.3 1

s REQUEST FOR INFORMATION (RI)

RI NO: _016 SYSTEM CODE: NRW ISSUE DATE: 7/31/87

SUBJECT:

MINIMUM SPRAY POND LEVELS DEPARTMENT: NED COORDINATOR: RON LAWRENCE TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /0UESTION:

The low level alarm setpoint for the Nuclear Raw Water Spray Ponds is below the minimum level required to met the seven day, post LOCA, criteria.

Process standards states that the level switch which causes the annunciator to sound in the Control Room is set for 5'7" + 1"/-3". This means the alarm could occur anywhere between 5'4" and 5'8". The purpose of this alarm is to indicate that the volume in the pond is at the minimum level required to meet the System Design Bases for a 7 day period of operation, post LOCA, without makeup water. The alarm setpoint was established by Z-NRW-M0243. This document states that the minimum volume required to meet the 7 day criteria is 2.59 x million gal which calculates out to 5.6' or 5'7.2". The minimum volume necessary to meet the 7 day criteria is calculated Z-NRW-H0245. It i states that the minimum volume required is 2.745 x million gal which I calculates out to about 5.72' or 5'8.64".

In summary, at present the spray pond levels can go below the minimum volume required to meet the 7 day, post LOCA scenario.

l ATTACHMENT 6.3 I

REQUEST FOR INFORMATION (RI)

RI NO: _ulf SYSTEM CODE: NRW ISSUE DATE: 7/31/87

SUBJECT:

ROOT CAUSE ANALYSIS OF LUBE OIL COOLERS PLUGGING DEPARTMENT: - NED COORDINATOR: RON LAWRENCE TEAM LEADER: M. AKINS POTENTIAL CONCERN /0UESTION:

A root cause analysis has not been performed to determine the cause of the plugging of tube oil coolers for safety related components.

• SSR Problem 6 addresses bearing temperatures trend analyses for the High Pressure Injection Pump, Makeup Pumps, Decay Heat Pumps and Reactor Building spray pumps. This problem was caused by plugging of lube oil coolers thus, reducing cooling water flow to the coolers.

The problem resolution is to implement a trending program using temperature readings on bearing cooling water inlet, not outlet.

• SSR Problem 7 addresses monitoring heat exchanger performance on NSRH cooled heat exchangers. This was recommended because of heat exchangers plugging.

The resolution for this problem is to modify the surveillance procedure i to obtain data to trend heat exchangers performance via pressure drop, not flow reduction. However, no acceptance criteria is addressed in the new surveillance procedure.

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. SSR Problem 8 addresses a program to measure tube side pressure drops on !

excnangers with tubeside NSRH cooling. The purpose of this is to measure partial flow blockage.

The resolution provided for this is to install test connections ana revise surveillance procedure to acquire cath necessary to trend neat f exchanger pressure drop, not flow rate reduction. Again, no acceptance criterion has been established.

Each of these SSR problems address the same thing, plugging. No documentation could be produced where a root cause analysis was done nr is being performert to determine the cause of the plugging. The solutions to the SSR problems do not resolve the problems, nor do they address the fluid dynamics of the systert, which indicate that the present flow rate may not be  ;

adequate with no plugging or fouling. They only address provisions to monitor the plugging / fouling happen. This does not ensure that the system will meet or be ready to meet its intended function during an unexpected incident.

ATTACHMENT 6.3 l