ML20237A449

Forwards Technical Evaluation Rept EGG-EA-6941, Conformance to Reg Guide 1.97 Davis-Besse Nuclear Power Station,Unit 1, Inadvertently Omitted from

ML20237A449
Person / Time
Site:	Davis Besse
Issue date:	12/04/1987
From:	De Agazio A Office of Nuclear Reactor Regulation
To:	Shelton D TOLEDO EDISON CO.
References
RTR-REGGD-01.097, RTR-REGGD-1.097 DB-87-020, DB-87-20, TAC-51084, TAC-51085, NUDOCS 8712150009
Download: ML20237A449 (29)

Text

- _ _ _ _ - _ _ _ _ .

a , December 4, 1987

• Docket No. 50-346 Serial No. 0B-87-020 Mr. Donald C Shelton Vice President, Nuclear Toledo Edison Company Edison Plaza - Stop 712 ,

300 Madison Avenue i Toledo, Ohio 43652

SUBJECT:

REVIEW 0F DAVIS-BESSE CONFORMANCE TO REGULATORY GUIDE 1.97, REVISION 3 (TAC 51085)

My letter of November 25, 1987, forwarded a Safety Evaluation relating to the staff's review of your submittals regarding conformance with Regulatory Guide 1.97. That Safety Evaluation refers to the enclosed Technical Evaluation Report, prepared by our consultant, which inadvertently was omitted.

Sincerely,  ;

oriainal sioned by l

Albert W. De Agazio, Project Manager Project Directorate III-1 Division of Reactor Projects - III, IV, V & Special Projects

Enclosure:

DISTRIBUTION Technical Evaluation Report 'Descket File :

' PD31" Plant Gray cc w/ enclosure: GHolahan See next page RIngram ADeAgazio 0GC-Beth EJordan JPartlow ACRS (10) l NRC & Local PDRs PM/PD31:0 SP D/PD31:DRSP/#

l ADeAgazio:1t MVirgilio 12/3 87

/ 12/3 /87 B712150009 973904 PDR P ADOCK 05000346 PDR t___________________________________-.____ _ _ _ _ _ _ _ _ __ _ __

ENCLOSURE EGG-EA-6941 l

1 I

l TECHNICAL EVALUATION REPORT CONFORMANCE TO REGULATORY GUIDE 1.97 DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 Docket No, 50-346 l

Alan C. Udy Published November 1986 l

Idaho National Engineering Laboratory EG&G idaho, Inc.

Idaho falls, Idaho 83415 l o l

) Q l'I\

v ' '

ggp;g g:-

Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 '

Under 00E Contract No. DE-AC07-76ID01570 FIN No. A6483 l

1 l

l w_-____-_______

I I

l l

I ABSTRACT l

This EG&G Idaho, Inc., report provides a review of the submittal for Regulatory Guide 1.97, Revision 3, for Unit No. I of the Davis-Besse Nuclear Power Station. Any exception to the guidelines of Regulatory Guide 1.97 are evaluated and those areas where sufficient basis for acceptability is not provided are identified.

l 1

I l

l l

l l

l Docket No. 50-346 i TAC No. 51084 l

l

)i l

lI l

l l

4 FOREWORD l This report is supplied as part of the " Program for Evaluating l

Licensee / Applicant Conformance to RG 1.97," being conducted for the l

U.S. Nuclear Regulatory Connission, Of fice of Nuclear Reactor Regulation, I Division of PWR Licensing-A, by EG&G Idaho, Inc., NRR and I&E Support Branch.

The U.S. Nuclear Regulatory Connission funded the work under

]

authorization B&R 20-19-10-11-3. l l

l l

l l

l l

l l

l l

Docket No. 50-346 TAC No. 51084 111

~

I' CONTENTS 1

ABSTRACT . . . . . . . . . . . ......... . ... ..... . .... 11 i J

FOREWORD ..... ... ... ............. ....... .... ... .... ........... iii

1. INTRODUCTION .. . . .. . . . . . .. ........ ...... ..... ............ l

2. REVIEW REQUIREMENTS . . ... ................. . . ...... ...... 2

3. EVALUATION .................. ................... ................ 4 3.1 Adherence to Regulatory Guide 1.97 ......................... 4 i 3.2 Type A Variables ........................................... 4 1

3.3 Exceptions to Regulatory Guide 1.97 .... ................... 5

4. CONCLUSIONS ................................ .. .................. 22 S. REFERENCES ........ . . .. . ... ...................... ............. 23 1

i i

l l

l 9

iv

CONFORMANCE TO REGULATORY GUIDE 1.97 DAVIS-BESSE NUCLFAR POWER STATION, UNIT NO. 1

1. INTRODUCTION On December 17, 1982, Generic Letter No. 82-33 (Reference 1) was issued by D. G. Eisenhut, Director of the Division of Licensing, Nuclear I Reactor Regulation, to all licensees of operating reactors, applicants for operating licenses, and holders of construction permits. This letter included additional clarification regarding Regulatory Guide 1.97, Revision 2 (Reference 2), relating to the requirements for emergency response capability. These requirements have been published as Supplement No. 1 to i NUREG-0737, "THI Action Plan Requirements" (Reference 3).

i The Toledo Edison Company, the licensee for the Davis-Besse Nuclear '

Power Station, provided a response to Section 6.2 of the generic letter on June 28, 1984 (Reference 4). This response provides a comparison of the licensee's instrumentation to the recommendations of Revision 3 of Regulatory Guide 1.97 (Reference 5). Additional inf ormation was submitted by the licensee on January 27, 1986 (Reference 6) and on September 26, 1986 (Reference 7).

1 This report provides an evaluation of this material.

i l

l 9

1

2. REVIEW REQUIREMENTS Section 6.2 of NUREG-0737, Supplement No. 1, sets forth the documentation to be submitted in a report to the NRC describing how the licensee complies with Regulatory Guide 1.97 as applied to emergency response facilities. The submittal should include documentation that provides the following information for each variable shown in the applicable table of Regulatory Guide 1.97.

1. Instrument range

2. Environmental qualification

3. Seismic qualification

4. Quality assurance

5. Redundance and sensor location

6. Power supply

7. Location of display

8. Schedule of installation or upgrade 4 l

The submittal should identify deviations from the regulatory guide and provide supporting justification or alternatives.

i Subsequent to the issuance of the generic letter, the NRC held regional meetings in February and March 1983, to answer licensee and applicant questions and concerns regarding the NRC policy on this subject.

At these meetings, tt was noted that the NRC review would only address exceptions taken to Regulatory Guide 1.97. Where licensees or applicants explicitly state that instrument systems conform to the regulatory guide, it was noted that no further staff review would be necessary. Therefore, 2

}

this report only addresses exceptions to Regulatory Guide 1.97. The following evaluation is an audit of the licensee's submittals based on the review policy described in the NRC regional meetings.

3

3. EVALUATION The licensee provided a response to item 6.2 of NRC Generic Letter 82-33 on June 28, 1984 Additional information was provided on January 27, 1986 and on September 26, 1986. These responses describe the licensee's I

position on post-accident monitoring instrumentation. This evaluation is based on that material.

3.1 Adherence to Regulatory Guide 1.97 l The licensee has provided a review of their post-accident monitoring ,

instrumentation that compares the instrumentation characteristics against Regulatory Guide 1.97, Revision 3. The licensee has provided a listing of the regulatory guide variables, wherein are listed compliance, deviations, and references to justification for any deviations. The licensee states that modifications identified to bring about compliance with the regulatory l guide will be accomplished in accordance with the licensee's Integrated Implementation Plan. Therefore, we conclude that the licensee has provided an explicit commitment on conformance to Regulatory Guide 1.97, except for those deviations that were justified by the licensee as noted in Section 3.3.

3.2 Type A Variables Regulatory Guide 1.97 does not specifically identify Type A variables, i.e., those variables that provide the information required for operator controlled safety actions. The licensee classifies the following instrumentation as Type A.

1. Reactor coolant system (RCS) hot leg water temperature l

2. RCS pressore i

3. Containment pressure 4

I

4. Containment hydrogen concentration

5. Residual heat removal system flow

6. Flow in high pressure injection system

7. Flow in low pressure injection system

8. Refueling water storage tank level

9. Pressurizer level

10. Steam generator pressure

11. Control room normal ventilation isolation status This instrumentation meets the Category 1 recommendations consistent with the requirements for Type A variables.

3.3 Exceptions to Regulatory Guide 1.97 1

The licensee identified exceptions to and deviations from Regulatory i

Guide 1.97. These are discussed in the following paragraphs.

3.3.1 Reactor Coolant System (RCS) Soluble Baron Concentration Regulatory Guide 1.97 recommends on-line instrumentation with a range of 0 to 6000 ppm to verify boron injection and reactivity control in a post-accident situation. The licensee has not provided this on-line instrumentation, but can obtain the information by utilizing the i post-accident sampling system and on-site laboratory analysis.

The licensee deviates from Regulatory Guide 1.97 with respect to post-accident sampling capability. This deviation goes beyond the scope of ,

5

i

~

this review and is being addressed by the NRC as part of their review of NUREG-0737, Item 11.B.3. l 1

)

3.3.2 RCS Cold Leg Water Temperature Regulatory Guide 1.97 recommends Category 1 instrumentation with a range of 50 to 700*F for this variable. The licensee has supplied Category 3 instrumentation with a range of 50 to 650*F. There is a deviation in both range and category.

The licensee states that the cold leg temperature will be less than 650*F for any design basis event. As the instrumentation will remain on scale for any anticipated event, we find the range of this instrumentation acceptable.

l The licensee's justification for Category 3 instrumentation follows.

l l

1. The cold leg temperature is not required to establish or verify natural circulation. If the RCS hot leg temperature and the core exit temperature when compared to the RCS pressure establishes that a subcooled condition exists and at least one steam j generator has established auxiliary feedwater flow as indicated by steam generator level, then natural circulation will be assured. It is due to this reasoning that the licensee has l identified RCS hot leg temperature, RCS pressure, and steam generator pressure as Type A variables.

2. The licensee states that they do not have to monitor cold leg temperature in order to prevent thermal shock to the reacter vessel due to excessive high pressure coolant injection flow (throttling), because the reactor vessel does not have high copper coctent or axial welds.

As the licensee has supplied Category 1 instrumentation for the variables RCS hot leg temperature, core exit temperature, steam generator level, and auxiliary feedwater flow, and because the licensee indicates l 6

l that the thermal shock is not a consideration, we find this justification for Category 3 RCS cold leg water temperature instrumentation acceptable.

3.3.3 RCS Hot Leg Water Temperature Regulatory Guide 1.97 recommends instrumentation with a range of 0 to 700*f for this variable. The licensee has supplied instrumentation with a range of 120 to 920*f. The licensee states that at temperatures less than 280*f, the decay heat removal system is used instead of the steam generators to cool the RCS. This system has additional temperature instrumentation to monitor the RCS in this temperature range. Category 1 .

1 core exit thermocouple also provide inf ormation below 120*f. 4 Based on the licensee's justification, we find that the range of 120 to 920*f is satisfactory.

3.3.4 RCS Pressure Regulatory Guide 1.97 recommends instrumentation for this variable with a range from 0 to 3000 psig. The licensee's existing instrumentation has a range from 0 to 2500 psig. In Reference 6, the licensee commits to extend the range of this instrumentation to comply with Regulatory '

Guide 1.97.. We find this commitment acceptable.

I 3.3.5 Radiation Level in Circulating Primary Coolant '

The licensee indicates that radiation level measurements to indicate fuel cladding failure are provided by the following instruments.

1. Letdown line radiation monitors

2. Post-acciqent sampling systetii 7

l

1

~

l l

The post-accident sampling system is available with the reactor isolated, and is being reviewed by the NRC as part of their review of NUREG-0737 Item 11.8.3.

1 Based on the alternate instrumentation provided by the licensee, we l conclude that the instrumentation supplied for this variable is adequate and therefore. acceptable. '

3.3.6 Analysis of Primary Coolant The licensee states that this variable is not necessary at the l Davis-Besse Nuclear Power Station.

l The licensee deviates from Regulatory Guide 1.97 with respect to post-accident sampling capability. This deviation goes beyond the scope of this review and is being addressed by the NRC as part of their review of NUREG-0737 Item II.B.3.

i l

3.3.7 Containment Effluent Radioactivity--Noble Gases from Identified l Release Points Effluent Radioactivity--Noble Gases (from Buldinos or i

Areas Where Penetrations and Hatches Are located) Radiation Exposure Rate (Inside Buildings or Areas Where Access is Required)

Reference 6 provided clarification on the information for these variaDies, for the first two, the licensee has identified the noble gas monitor for the common plant vent for monitoring these variables. This instrumentation exceeds the recommendations for the instrumentation specified by the regulatory guide for these variables and is acceptable.

The instrumentation for the variable radiation exposure rate has a range that goes up to 10 R/hr rather than 10 R/hr as recommended by the regulatory guide. The licensee states that the instrumentation has a range that is sufficient to provide personnel protection and alert the operator 8

I to the potential of high radiation fields. This instrumentation is supplemented by portable instrumentation.

from a radiological standpoint, if the radiation levels reach or exceed the upper limit of the range, personnel would not be permitted into the areas without portable monitoring (except for life saving). Based on the alternate supplemental instrumentation used by the licensee for this variable, we find the range for the radiation exposure rate monitors acceptable.

3.3.8 Residual Heat Removal (RHR) Heat Exchanger Outlet Temperature Regulatory Guide 1.97 recorrsnends Category 2 instrumentation for this variaole. The licensee has provided instrumentation that exceeds Category 3 requirements but that is not Category 2, to monitor both the inlet and the outlet temperatures of the heat exchangers. The licensee states that Category 3 instrumentation is acceptable because the safety features actuation system signals cause maximum cooling to take place, via opening fully the component cooling water heat exchanger service water outlet valves and the decay heat removal coolers component cooling water outlet valves. Maximum decay heat removal occurs without operator intervention.

The operator monitors the performance of the decay heat removal system by observing the Category 1 instrumentation for the decay heat removal system flow and the core exit temperature.

Based on the system flow and core exit temperature being the key variables to monitor the decay heat removal system, we find the instrumentation provided for the decay heat rerr. oval heat exchanger outlet temperature acceptable.

3.3.9 Accumulator Tank Level and Pressure Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable with a range of 10 to 90 percent volume and 0 to 750 psig. The 9

l licensee has identified deviations in this instrumentation: a range of 0 to 14 feet (2.4 to 84.1 percent) level and Category 3 pressure instrumentation with a range of 0 to 700 psig.

I l

The licensee states that the level is verified and manually controlled to between 7555 and 8004 gallons of borated water. The upper limit of this technical specification requirement is less than 76 percent of the tank volume, and withir. the limits of the range supplied. Therefore, we find the level range provided for this instrumentation acceptable.

l Table 6.11 of the Final Safety Analysis Report (FSAR, Reference 8) indicates that the accumulator pressure is manually controlled at 600 115 psig. In addition, there are relief valves that relieve pressure in excess of 700 psig, the tank design pressure. Therefore, we find the range of the pressure instrumentation acceptable. 1 The accumulators 4re passive, and the licensee indicates that there are no manual actions taken as a result of the pressure indication. All operator actions are based on pressurizer level. Thus, the licensee concludes that Category 3 instruments are adequate for the backup instrumentation used for the accumulator pressure. '

We find that the 0 to 700 psig, Category 3 instrumentation is satisfactory for the pressure portion of this variable.

3.3.10 Boric Acid Charging Flow Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable. The licensee has Category 3 instrumentation. The licensee states that this variable is not part of a safety-related system. It is the boration pathway for normal operation. For post-accident situations the borated water storage tank provides boration via the high and the low pressure coolant injection.

10

Based on the information provided by the licensee, we find the deviation from Category 2 to Category 3 for this variable acceptable.

3.3.11 Pressurizer Level Regulatory Guide 1.97 recommends instrumentation for this variable with a range from the top to the bottom of the pressurizer vessel. The licensee states that the range of this instrumentation is 0 to 320 inches (or G.7 to 78.2 percent of the pressurizer volume).

a l

The licensee states that only severe transients or accidents will l cause the pressurizer level instrumentation to go off scale either by voiding the pressurizer or by the reactor coolant system (RCS) becoming solid. The lower pressurizer tap is as close to the bottora of the pressurizer as is reasonable. Voiding of the pressurizer is observed by the rapid decrease of the RCS pressure following the loss of pressurizer level indication and resultant changes in the subcooled mar gin. Indication of the RCS becoming water solid is observed by a high RCS pressure, accompanied by possibly large fluctuations in pressure, power operated relief valve position and pressurizer safety valve position indication.

These are monitored by Category 1 instrumentation in the control room. >

Based on the licensee's justification, we find the range provided for the pressurizer level instrumentation acceptable.

3.3.12 Pressurizer Heater Status I

Regulatory Guide 1.97 recomends instrumentation to monitor the

{

current drawn by the pressurizer heaters. The licensee's control room instrumentation consists of on/off indication of the redundant emergency pressurizer heaters (the licensee indicates that the control of these heater banks is eith,er on or of f, and theref ore the instrumentation is appropriate); diesel generator voltmeters, ammeters and wattmeters; 4160 V essential bus voltmeters and ammeters; voltmeters and ammeters on the 4160/480 V transformers and 480 V bus voltmeters. Thus, instrumentation to l

11 i I

______________________ _ _ _ - _ _ __ I

a preclude overloading an emergency power source is provided. We find this alternate instrumentation acceptable.

3.3.13 Ouench Tank Level Regulatory Guide 1.97 recommends instrumentation for this variable with a range from the top to the bottom of the tank. The range supplied by the licensee indicates the straight cylindrical shell height, excluding the hemispherical ends of the tank.

The portion of the quench tank level that is not indicated is in the upper and lower hemispherical head regions, where the volume to level ratio is not linear. This an acceptable deviation from Regulatory Guide 1.97.

3.3.14 Quench Tank Temperature Regulatory Guide 1.97 recommends instrumentation for this variable with a range from 50 to 750*F. The installed instrumentation has a range of 0 to 400*F.

The licensee states that the quench tank rupture disc relieves any I pressure in excess of 100 psig. The saturated steam at this pressure will not exceed 338*F.

We find the licensee's justification for this deviation acceptable.

3.3.15 Steam Generator Level Regulatory Guide 1.97 recommends Category 1 instrumentation with a range from tube sheet to separators. This is for U-tube steam generators.

The Davis-Besse steam generators are of once-through design, and as such the heat exchange area would be described as tube sheet to tube sheet.

The licensee's steam generator level instrumentaticq consists of a) one full range channel (0 to 641 inches), b) two operating channels (96 to 388 inches), c) two startup channels (0 to 250 inches) that are Category 1 12

i 1

I and d) four startup channels (0 to 388 inches) that are Category 1. The zero reference is at 6 inches above the lower tube sheet. Should the water level reach this point, the steam generator is essentially dry.

The licensee has reviewed the design basis accidents and found that the full range instrumentation is not used by any abnormal or emergency operating procedure. Thus, it is not required in the mitigation of any ]

design basis accident. The full range instrumentation is used when the reactor is shutdown and the steam generators are in a wet lay-up condition. It is also used as backup instrumentation for the operating and l startup instrumentation. As backup instrumentation Category 3 instrumentation is acceptable. j The operating channels do not initiate any automatic safety function.

Although they are used for a steam generator tube rupture or steam generator overfill, these events do not create a harsh environment, and the Category i startup channels can be used by the operator over the same range. Therefore, we find Category 3 instrumentation for the operating channels acceptable.

l l

The licensee states that, for those analyzed transients and accidents that make use of the Category 1 startup instrumentation, tne instrumentation will remain f unctional and on scale. Based on this statement, we find the range provided for the Category 1 startup channels acceptable.

3.3.16 Steam Generator Pressure Regulatory Guide 1.97 recommends instrumentation for this variable with a range from 0 to 20 percent above the lowest safety valve setting.

The lowest safety valve setting is 1050 psig; therefore, the range should be from 0 to 1260 ps,ig. The instrumentation for this variable has a range ]

of 0 to 1200 psig, 9 percent above the highest safety valve setting.

l 1

The licensee states that the pressure indication will remain on scale, including the most severe design basis accident. Based on this statement, 13 l

and the maximum range being 100 psi above the highest safety valve setting, we find that the range of 0 to 1200 psig is acceptable.

3.3.17 Safety / Relief Valve Positions or Main Steam Flow Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable. The licensee states, in Reference 6, that Category 2 position indication will be provided for the atmospheric vent valves (AVV) and the main steam safety valves (MSSV). We find this commitment acceptable.

j i

3.3.18 Condensate Storage Tank Water Level J l

Regulatory Guide 1.97 recommends Category 1 instrumentation for this variable. The licensee has Category 3 instrumentation. The nuclear safety related feedwater supply is from the service water (SW) system. This is 4 the assured water supply. The switchover from tne condensate storage tank (CST) to SW is automatic. The licensee states that the switchover has been reviewed and approved by the NRC (see letter dated February 21, 1984, TED No. 1455, comment GS-4). The licensee states that the CST serves no nuclear safety-related function other than being the preferred water source. Therefore, the licensee concludes that Category 3 instrumentation 1 is appropriate for this variable. I i

)

We find this to be a good feith attempt, as defined in NUREG-0737, Supplement No. 1, Section 3.7 (Reference 3), to meet NRC requirements and I is, therefore, acceptable.

3.3.19 Containment Spray Flow Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable.

In addition to Category 3 flow instrumentation, the licensee has alternate Category 1 instrumentation, consisting of containment spray pump on indication and containment spray valve position indication. The containment spray valves are actuated automatically on either a decrease in reactor coolant system pressure or an increase in containment pressure.

14

The containment spray pumps are actuated automatically on a further increase in containment pressure. Thus, the licensee considers the Category 1 containment pressure instrumentation as the key variable in ,

l verifying containment integrity, and the containment spray flow indication i as a backup variable. The effectiveness of the containment spray can be shown by the containment temperature and pressure trends.

The instrumentation provided by the licensee is adequate to monitor this variable. Therefore, we find this instrumentation acceptable.

3.3.20 Containment Atmosphere Temperature

! Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable with a range from 40 to 400*f. The licensee has supplied i j instrumentation with a range of 0 to 300*f. The instrumentation is l

Categcry 2 except in the brea of environmental qualification. The NRC i l

l exempted it from the requirements of 10 CFR 50.49. We find tnis acceptable. '

I l The licensee indicates that the maximum containment temperature will be less than 285"F. Therefore, the range of 0 to 300*f is acceptable, and the instrumentation for this variable is acceptable.

3.3.21 Containment Sump Water Temperature Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable with a range of 50 to 250*f. The licensee does not have direct l instrumentation for this variable. Their justification is that monitoring 1

the sump temperature is not needed to assure that net positive suction head 1

(NPSH) exists for the decay heat pumps or the containment spray pumps.

They state that containment sump water temperature is not required to mitigate the consequences of any design basis accidents.

9 The licensee has shown that the safety-related core and containment cooling system pumps will have adequate net positive suction regardless of sump water temperature. The licensee establishes that containment cooling is taking place with a variety of instrumentation, including: decay heat 15

removal (DHR) heat exchanger inlet and outlet temperatures (Category 3),

DHR system flow (Category 1), containment spray flow (Category 3),

containment spray pump and valve position indication (Category 1),

containment pressure (Category 1), containment atmosphere temperature (Category 3), reactor coolant system hot and cold leg water temperatures (Category 1) and core exit temperature (Category 1).

With adequate net positive suction head provided and suitable alternative instrumentation to show the effectiveness of the containment and core cooling, we find this deviation acceptable.

l 3.3.22 Makeup Flow-In i

Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable with a range of 0 to 110 percent of design flow. The licensee indicates, in Reference 6 that their instrumentation exceeds this range, covering from 0 to 160 gpm (0 to 115 percent of design flow).

The licensee states that this indication goes off-scale when a second makeup pump is started to maintain the pressurizer level. Because of this, the licensee has committed to provide Category 2 instrumentation with a range that will meet or exceed 110 percent of the design flow of both nakeup pumps. We find this commitment acceptable.

3.3.23 Letdown Flow-Out Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable. The licensee has provided instrumentation that, except for environmental qualification, is Category 2. The licensee states that this.

variable is not required in the mitigation of an accident, and that the letdown system is isolated by accidents requiring containment isolation.

The licensee also states that this instrumentation has been addressed by the Environmental Qualification Rule, 10 CFR 50.49, and found exempt.

Based on this, we find the instrumentation provided by the licensee for this variable acceptable.

16

l l

3.3.24 Volume Control Tank Level  !

Regula tory Guide 1.97 reccamends ins trumenta t ion f or this variable with a range from the top to the bottom of the tank. The licensee does not consider this as post-accident instrumentation; however, the range supplied by the licensee indicates the straight cylindrical shell height, excluding the hemispherical ends of the tank. Also the licensee indicates that this tank is not required in an accident situation, as system isolation occurs.

Based on the licensee's justification for not requiring this instrumentation in a post-accident situation, we find this deviation in range acceptable.

i l

3.3.25 Component Cooling Water (CCW) Temperature to Engineered Safety features (ESf) System Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable. The licensee is supplying Category 3 instrumentation. Two of the three component cooling water heat exchanger outlet temperatures are j monitored in the control room.

1 i

l During a design basis accident, upon actuation of the Safety features Actuation System (SFAS), Incident Level 2, the service water inlet valves to the CCW heat exchangers are opened providing maximum cooling water to the heat exchanger. The licensee states that no actions are required based upon this temperature and the valve positions are indicated in the control room and the valves open automatically by safety grade control systems.

For these reasons, the licensee considers this instrumentation as Category 3 backup instrumentation.

We find the justification for the existing instrumentation for this variable acceptable, 17

3.3.26 Component Coolinq Water (CCW) Flow to ESF System Regulatory Guide 1.97 recommends Category 2 flow instrumentation to monitor the operation of the component cooling water system. The licensee has alternate control room instrumentation for this variable, consisting of the following:

a) CCW pump motor indication (Category 1) b) CCW system valve position indication (Category 1) c) system / equipment'using CCW flow

1) high temperature alarm and indication

2) low CCW flow alarm (local indicator) d) CCW surge tank level indication and alarms I

l The licensee uses the above instrumentation to monitor the operation of the CCW system. The licensee states that the design flow to the various safety-related components varles with the equipment. The licensee states j that, due to the wide range of design flows to these components, total loop flow would not be indicative of overall system performance.

1 Based on the justification provided by the licensee, we find the a? ternate instrumentation provided acceptable.

3.3.27 High level Radioactive Liquid Tank Level l Regulatory Guide 1.97 recommends instrumentation for this variable with a range from the top to the bottom of the tank. The range supplied by the licensee indicates the straight cylindrical shell height, excluding the hemispherical ends.

18 l

l

The portion of the high level radioactive liquid tank level that is not indicated is in the upper and lower hemispherical head regions, where the volume to level ratio is not linear. This is an acceptable deviation from Regulatory Guide 1.97.

l 3.3.28 Radioactive Gas Holdup Tank Pressure Regulatory Guide 1.97 recommends instrumentation for this variable with a range of 0 to 150 percent of design pressure. The tank design pressure is 150 psig and pressure above this is automatically relieved.

The instrumentation range is 0 to 200 psig. Thus, the range deviates from j

the recommended range. i There is a safety relief valve on this tank, set to relieve any pressure above 150 psig. An alarm is set at 140 psig. As the tank pressure will not exceed 150 psig, we find the 0 to 200 psig range )

acceptable.

3.3.29 Emergency Ventilation Dameer position

]

Regulatory Guide 1.97 recommends monitoring the open-closed status of these dampers. The licensee states that all emergency ventilation system (EVS) dampers have the recot:nended indication except for the f an inlet dampers which do not have indication in the control room but are .

l interlocked to open when the auxiliary building normal ventilation dampers '

are closed. Differential pressure across the filter banks is provided in the control room. Overall EVS system performance can be determined from the annulus to mechanical penetration Category 1 differential pressure indicators, located in the control room.

We find the licensee's alternate instrumentation for this variable acceptable. ,

i 19

. 3.3.30 Noble Gas Vent from Steam Generator Safety Relief Valves or Atmospheric Dump Valves Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable. The recommended parameters to be monitored for this variable are noble gas (10 to 10 pCi/cc), duration of release in seconds and mass of steam per unit time.

The licensee's instrumentation has a range of 10 to 106counts per minute (CPM). When used to detect N-16 (the earliest reactor coolant radioisotope detectable), this range is equivalent to 1.1 x 10- to 1.1 x 10~ pC1/cc. For release assessment (the purpose as stated by the regulatory guide), the range is equivalent to 9 x 10-5 to 9 uCi/cc at the start of an accident. Due to the attenuation of the steamline wall thickness and shielding, the sensitivity (CPM per pCi/cc) as a function of time and the source volume, eight hours after the start of an accident, the range is equivalent to 10-3 to 102 pCi/cc.

l The licensee as calculated the radioactive nuclide content for a steam '

ger.erator tube rupture for which this instrumentation is used. In the initial accident phase, the level would be 2.4 x 30~ pCi/cc (within the range provided at this time frame). Worst case calculations put'an upper bound on the content at 2.2 pCi/cc (also within the range pruvided). l The instrumentation and been addressed in accordance with 10 CFR 50.49 (environmental qualification) and found to be acceptable for-their environment when they are needed. The licensee is also installing instrumentation for the position of the atmospheric vent valves and the main steam safety valves, (see Section 3.3.17). This, in conjunction with the steam generator outlet pressure, will allow the calculation of time of release and mass of steam per unit time.

Based on the licensee's justification and procedures, we find the instrumentation supplied for this variable acceptable.

20

3.3.31 Estimation of Atmospheric Stability i Regulatory Guide 1.97 recommends instrumentation for this variable with a range of -9 to +18*f or an analogous range for alternative stability 1

analysis. The licensee has supplied instrumentation with a range of 4 to

+8'F. The licensee justifies this, indicating that the range is based on the Pasquill Stability Class specified in Regulatory Guide 1.23.

i Table 1 of Regulatory Guide 1.23 provides seven atmospheric stability J classifications based on the difference in temperature per 100 meters elevation change. These classifications range from extremely unstable to extremely stable. Any temperature difference greater than +4*C or less

• 1 than -2*C does nothing to the stability classification. The licensee's instrumentation includes this range. Therefore, we find that this i

instrumentation is acceptable to determine the atmospheric stability. I f

i 3.3.32 Accident Sampling (Primary Coolant. Containment Air and Sump) j l l The licensee's post-accident sampling system provides sampling and analysis as recommended by the regulatory guide, except that l l

1. The emergency core cooling system pump room sump is not sampled l l

l

2. It does not have containment air oxygen content analysis {

3. Containment air hydrogen and gamma spectrum analysis are done by 1 on-line monitors in preference to grab samples.

The licensee deviates from Regulatory Guide 1.97 with respect to post-accident sampling capability. This deviation goes beyond the scope of this review and is being addressed by the NRC as part of their review of NUREG-0737, Item 11,B.3.

1 5

21

i I

. 4. CONCLUSIONS Based on our review, we find that the licensee either conforms to or is justified in deviating from Regulatory Guide 1.97..

l l

l l

l l

l

+

22

q S. REFERENCES

1. NRC letter, D. G. Eisenhut to All Licensees of Operating Reactors, Applicants f or Operating Licenses, and Holders of Cons truction Permits, " Supplement No. I to NUREG-0737--Requirements for Emergency Response Capability (Generic Letter No. 82-33)," December 17, 1982.

2. Instrumentation for Liqht Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions Durinq and f ollowing an Accident, Regulatory Guide 1.97, Revision 2. U.S. Nuclear Regulatory Commission (NRC), Office of Standards Development, December 1980.

3. Clarification of THI Action Plan Requirements, Requirements for l

Emergency Response Capability, NUREG-0737, Supplement No. 1, NRC, Office of Nuclear Reactor Regulation, January 1983.

4. Toledo Edison Company letter, R. P. Crouse to Director of Nuclear Reactor Regulation, NRC, June 28, 1984, Serial No. 1059. '

i S. Instrumentation for light Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Followinq an Accident,

)

l Regulatory Guide 1.97, Revision 3, NRC, Office of Nuclear Regulatory Research, May 1983.

6. Toledo Edison Company letter, J. Williams, Jr. to J. F. Stolz, NRC l January 27, 1986, Serial No. 1232. I

7. Toledo Edison Corapany letter, J. Williams, Jr. to J. F. Stolz, NRC, September 26., 1966, Serial No. 1266.

8. Final Safety Analysis Report, Davis-Besse Nuclear Power Station, Unit No. 1, Toledo Edison Company.

l l

• l 23

y 1 Nuckgam e tcyLafon v COWue13,04 I

• t PC# 7 %v*914 +da t'*s ew f 4C sad ved No , .e de.v a a.aC.,80AM

,s t 339 7".,',"3 Z2 BIBLIOGRAPHIC DATA SHEET EGG-EA-6941

$tt saettavCtlON5 0% T-t as visst 2 7 vt t a%Q $ce r. igg Jklavt9 LANE Conformance to Regulatory Guide 1.97, Davis-Besse fluclear Power Station, Unit tio. 1 4 Oa't at>0m' covettitD vest *taa

,,,_Q,,,, tiovember 1986 Alan C. Udy "*""

ocs1 ,taa g

November 1986 1 #tm60mue%G CmGaqqario% haut Aaso wast.NG acontSS "wevse /* Co88, a seQAG7'f a$a WQpa 6 Net NWet A EG&G Idaho, Inc. , ,,,, ,,0,A,, , , , , , ,

P.O. Box 1625 Idaho Falls, ID 83415 A6483

.o s.o~so ,~o o.ca~qc.o~ sa.t a~o oa,s, o4 0 0 13 3 . ,,,.., < C , .. r n o. -l.ca1 Division of PWR Licensing-A Technical Evaluation Report Office of fluclear Reactor Regulation *""'*'"""~~'~*~

U. S. tiuclear Regulatory Commission Washington, DC 20555 22 SuP*LtwtNf am

• NOYt3

') ad3TR ACT GGo meres er *ean This EG&G Idaho, Inc. report reviews the submittals for the Davis-Besse .'luclear Power Station, Unit tio.1, and identifies areas  :

1 of non-conformance to Regulatory Guide 1.97. Exceptions to these I guidelines are evaluated.

l t o OCCVMt NT a% Ak v$,$ . a 4 4 rwC*CS.Cisca'P'O*3 ' S * * * ' 6 ' W Y '

, ETaftutNT Unlimited Distribution it steu.it.c .ass...cario~

. n,

, ,et~ri,,t.s,o,t ~ tset o n. s Unclassified

,r. u Unclassified a wo.l. c. .acu

. . .. ,c t

_ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _