ML18029A354

Conformance to Reg Guide 1.97,Browns Ferry Nuclear Plant, Units 1,2 & 3, Interim Rept

ML18029A354
Person / Time
Site:	Browns Ferry
Issue date:	12/31/1984
From:	Udy A EG&G, INC.
To:	NRC
Shared Package
ML18029A353	List: ML18029A354
References
CON-FIN-A-6483, RTR-REGGD-01.097, RTR-REGGD-1.097 TAC-51073, TAC-51074, TAC-51075, NUDOCS 8502040766
Download: ML18029A354 (28)

Text

INTERIM REPORT CONFORMANCE TO REGULATORY GUIDE'1.97 BROWNS FERRY NUCLEAR PLANT, UNIT NOS. 1, 2 AND 3 A. C. Udy Published December 1984 EGRG Idaho, Inc.

Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE, Contract No. DE-AC07-76ID01570 FIN No. A6483 85020407bb 850123 Foe AoaCX 05000259 I

F PDR

ABSTRACT This EGEST Idaho, Inc., report provides a review of the submittal for Regulatory Guide 1.97, Revision 3, for Unit Nos. 1, 2 and 3 of the Browns Ferry Nuclear Plant. Any exception to the guidelines of Regulatory Guide 1.97 are evaluated and those areas where sufficient basis for acceptability is not provided are also identified.

FOREWORD This report is supplied as part of the "Program for Evaluating Licensee/Applicant Conformance to RG 1.97," being conducted for the U.S.

Nuclear Regulatory Commission, Office of Nuclear reactor Regulation, Division of Systems Integration, by EGA Idaho, Inc., NRC Licensing Support Section.

The U.S. Nuclear Regulatory Commission funded the work under authorization BER 20-19-10-11-3.

Docket No. 50-259,50-26Q and 50-296 TAC Nos. 51073, 51074 and 51075

CONTENTS ABSTRACToeoo ~ ~ ~ o ~ ~ o ~ ~ o ~ ~ ~ o ~ oo ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ o ~ o ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~

F OREWORD................................................... ii

1. INTRODUCTION.......................................... 1

2. REVIEW REQUIREMENTS................................... 1

3. EVALUATION............................................ 2 3.1 Adherence to Regulatory Guide 1.97.... ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 0 2 3.2 Type A Variables...................... 0 ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ 3 3.3 Exceptions to Regulatory Guide 1.97... ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ 3

4. CONCLUSIONS................................ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 20

5. REFERENCES............................................ 23

CONFORMANCE TO REGULATORY GUIDE 1.97 BROWNS FERRY NUCLEAR PLANT UNIT NOS. 1 2 AND 3

1. INTRODUCTION On December 17, 1982, Generic Letter Ko. 82-33 (Reference 1) was issued by D. G. eisenhut, Director of the Division of Licensing, Nuclear 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 1 to NUREG-0737, "TMI Action Plan Requirements" (Reference 3).

Tennessee Valley Authority, the licensee for the Browns Ferry Nuclear Plant, provided a response to Section 6.2 of the generic letter on April 30, 1984 (Reference 4).

This report provides an evaluation of that material.

2. REVIEW REQUIREMENTS Section 6.2 of NUREG-0737, Supplement 1, sets forth the documentation to be submitted in a report to the NRC describing how the licensee meets the guidance of 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. Envir onmental 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.

Further, the submittal should identify deviations from the guidance in the regulatory guide and provide supporting justification or alternatives.

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 matter. At these meetings, it was noted that the NRC review would only address exceptions taken to the guidance of Regulatory Guide 1.97. Further, where licensees or applicants explicitly state that instrument systems conform to the provisions of the guide it was noted that no further staff review would be necessary.

t Therefore, this report only addresses exceptions to the guidance of 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. EVALUATION The licensee provided a response to Item 6.2 of the NRC generic letter 82-33 on April 30, 1984. The response describes the licensee's position on post-accident monitoring instrumentation. This evaluation is based on that material.

3.1 Adherence to Re ulator Guide 1.97 The licensee has provided a review of their post-accident monitoring instrumentation that compares the instrumentation characteristics against the recommendations of Requlatory Guide 1.97, Revision 3 (Reference 5).

The licensee provid'ed "a detailed evaluation of Regulatory Guide 1.97 requirements and implementation plans for the Browns Ferry Nuclear Plant" and our review only addresses the exceptions to these requirements.

Therefore, it is concluded that the licensee has provided an explicit commitment on conformance to the guidance of Regulatory Guide 1.97, except for those deviations that were )ustified by the licensee as noted in=Section 3.3.

3.2 T~AV I b1 Regulatory Guide 1.97 does not specifically identify Type A variables, i.e., those variables that provide information required to permit the control room operator to take specific manually controlled safety actions. The licensee classifies the following instrumentation to be their Type A variables.

1. Containment hydrogen concentration

2. Drywell pressure

3. Drywell atmosphere temperature All of the above variables are also included as Type B, C, or D variables.

3.3 Exce tions to Re ulator Guide 1.97 The licensee identified the following deviations from the recommendations of Regulatory Guide 1.97.

3.3.1 Oesi n Cate or Exce tions The following variables are identified as Category 2 by Regulatory Guide 1.97, while the licensee has furnished instrumentation for these variables which satisfy Category 3 requirements:

l

- Effluent radioactivity Noble gases Primary system safety relief valve position

- Reactor core isolation cooling flow

- High pressure coolant injection flow

- Core spray flow

- Low pressure coolant injection flow

- Standby liquid control system storage tank level

- Residual heat removal (RHR) system flow

- RHR heat exchanger outlet temperature

- Cooling water temperature toESF system components

- Cooling water flow to ESF system components

- Emergency ventilation damper position

- Status of standby power

- Noble gases and vent flow rate common plant vent.

The licensee indicates that these will meet or exceed Category 3 requirements, and that these variables are not needed for ensuring design basis behavior or for ma)or contingency actions. The licensee indicates that these are not essential and therefore Category 3 is adequate.

The exception of all the above variables from Category 2 recommendations is not acceptable. The licensee should, on a case .by case basis, identify the specific exceptions from the Category 2 recommendations and provide adequate justification for each exception.

3.3.2 Neutron Flux Regulatory Guide 1.97 recommends Category I instrumentation for this variable. The licensee has provided instrumentation which is not Category

1. The licensee states that "most portions of the neutron monitoring systems are designed, procured and tested to standards more stringent than Category 3." They identify the following components as not meeting Category I requirements: source and intermediate range monitors (SRM and IRM) drive mechanisms and controls and the neutron monitoring system power sources.

The licensee concludes that, due to historical reliability and the redundancy of the overlapping channels (4 SRM channels, 8 IRM channels and 6 average power range monitors), the currently installed instrumentation "meets the intent of the guide."

This deviation is similar to most BWRs. A Category 1 system that meets all the criteria of Regulatory Guide 1.97 is an industry development item.

Based on our review, we conclude that the existing instrumentation is acceptable for interim operation. The licensee should follow industry development of this equipment, evaluate newly developed equipment, and install Category 1 instrumentation when it becomes available.

3.3.3 Coolant Level in Reactor Regulatory Guide 1.97 recommends redundant Category 1 instrumentation for this variable with a range extending from the bottom of the core support plate to the centerline of the main steamline or the top of the vessel (whichever is less). The licensee has instrumentation that covers, with overlapping ranges, from I/3 of the core height to above the centerline of the main steamline.

Redundancy is provided up to 70 inches below the centerline of the main steamline. The shutdown vessel flooding range, which measures above this height, has no redundancy.

The shutdown vessel flooding range instrument reference leg uses the top 1

head vent as a penetration. In order to comply with the single failure requirement of Regulatory Guide 1.97, an additional head penetration would be needed for a redundant reference column for a second shutdown vessel flooding range channel.

Only the upper 70 in. of the recommended range are not monitored by redundant instruments. The licensee notes that no manual or automatic functions are initiated in the upper 70 in. since these functions occur in the range monitored by redundant channels. The licensee concludes that the reactor coolant level instrumentation meets the intent of the Regulatory Guide, and that only a marginal improvement in plant safety would be achieved by installing a redundant shutdown vessel flooding range channel. We find that an additional shutdown channel may result in only a marginal safety improvement and therefore find this justification acceptable.

The licensee did not provide justification for the lack of coverage of the lower 1/3 of core height. The licensee should either provide this justification or provide the recommended range.

The licensee states that "no recorder from a qualified instrument channel will be provided. This information is not essential for the operator 's direct and immediate trend or transient information. However, a level recorder from a nonqualified instrument loop is provided for both ranges."Item 6 of Table I of Revision 3 of Regulatory Guide 1.97 states the following. "If direct and

immediate trend or transient information is essential for operator information or action, the recording should be continuously available on redundant dedicated recorders. Otherwise, it may be continuously updated, stored in computer memory, and displayed on demand." Based on the licensee's statement that this information is not essential for the operator, we find that dedicated qualified recorders are not required for this variable.

3.3.4 Reactor Coolant S stem Pressure The licensee states that they will not provide a recorder for the Category 1 reactor coolant system pressure channels. The Justification provided by the licensee states that "this information is not essential for the operator's direct and immediate trend or transient information. It does, however, have a pressure recorder displayed in the control room that gets its signal from a nonqualified instrument loop that the operators use during normal operation. Reactor pressure will also be included in the database for the Safety Parameter Display (SPDS )."

As in the same deviation for the variable coolant level in reactor (Section 3.3.3), and based on the licensee's statement that "this information is not essential for the operator," we find the lack of dedicated qualified recorders are not required for this variable.

3.3.5 Or ell Sum Level Dr ell Grains Sum Level Regulatory Guide 1.97 recommends Category 1 instrumentation for these variables. The sumps at Browns Ferry use level switches to initiate sump pump out. Timers indicate the duration of sump pump operation for estimating the amounts of leakage. No safety-related system is actuated either automatically or manually as a result of the sump level. The drywell rump systems are automatically isolated at the primary containment penetration should an accident signal occur.

For small leaks, this Category 3 instrumentation will continue to function as the drywell temperature and pressure will not have changed

significantly. Therefore, the sump drains flow can be used as a leading indicator of reactor coolant system leakage. For larger leaks, the sumps will fill promptly, negating this instrumentation because the sump drain lines isolate due to the increase in drywell pressure caused by the accident. The sumps can be assumed full once containment isolation occurs at 2 psig.

In either case, we find the Category 3 instruments provided for this variable acceptable.

3.3.6 Primar Containment Pressure The licensee utilizes the instrumentation for the variable drywell pressure to measure this variable.

The licensee states the following. "The drywell and torus pressure may not be equal at all times; however, the drywell vacuum breakers have been designed to give assurance that torus pressure does not exceed drywell pressure by more than 2 psi. Also, the drywell pressure cannot exceed the torus pressure by greater that 1.5 psi due to the vent piping connections below the torus water level. Orywell/torus differential pressure is bounded by 1.5< d < 2 psi. Therefore, drywell pressure can be used as an indication of torus pressure and drywell pressure provides the key variable for monitoring primary containment'ressure."

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

3.3.7 Primar Contairment Isolation Valve Position Regulatory Guide 1.97 recommends Category 1 instrumentation for this variable. The licensee has listed those valves that comply with this position and those that do not. Those that are identified as not complying are grouped below, with the justification following.

There are three valves that are locked closed. The licensee does not indicate surveillance or administrative controls to assure these valves are closed. Therefore, we find this deviation not acceptable.

There are two valve pairs in the sampling and water quality system that are normally closed. The licensee states that a "local push button energizes two valves (A and B) simultaneously." The licensee does not indicate surveillance or administrative controls to assure these valves are closed.

Therefore, we find this deviation not acceptable.

The licensee should show that these seven valves are known to be closed in the post-accident situation.

1 The isolation valve on the high pressure coolant injection line is locked open. The licensee states that isolation is provided by a check valve. We find this deviation acceptable, as isolation is assured by the check valve.

There are 20 containment inerting system valves that are part of the hydrogen and oxygen monitoring system. These valves do not have position indication. They are par t of the post-accident monitoring instrumentation implemented for Regulatory Guide.l.97. Two control switches each control the position of half of these valves (they are grouped in 2 redundant systems).

Each system has its isolation signal indicated, which can be manually overridden by a manual key locked switch.

The purpose of this variable is to verify the accomplishment of the isolation function. Indicating the isolation signal does not verify that the signal actuated the valves, or that the individual valves have performed the isolation function. The licensee should provide the recommended instrumentation for these twenty valves.

3.3.8 Radiation Level in Circulatin Primar Coolant The licensee states that the critical actions to be taken to prevent and mitigate a gross breach of fuel cladding are (a) shut down the reactor, and (b) maintain water level. Neither of these are influenced by the recommended 10

variable. The applicant indicates that the post-accident sampling facility (PASF) provides a means of obtaining samples of reactor coolant and determining the status of fuel cladding and that radiation monitors in the condenser off-gas and main steamlines provide information on the status of fuel cladding when the plant is not isolated.

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

3.3.9 Primar Containment Area Radiation Hi h Ran e Regulatory Guide 1.97 recommends Category 1 instrumentation for this variable. The licensee has supplied Category 3 instrumentation, that meets some of the Category 1 requirements. The specific deviations are not identified nor is Justification provided. Therefore, we cannot conclude that the instrumentation supplied for this variable is acceptable.

The licensee should identify the specific deviations for this instrumentation, and provide justification for these deviations.

3.3.10 Containment and Dr ell Ox en concentration Regulatory Guide 1.97 recommends Category 1 instrumentation for this variable. The licensee has Category 3 instrumentation. The licensee provided the following to support this deviation. "The function of detection of a potential for a breach in the containment is also monitored by the following var i abl es;

1. Drywell pressure Type A, Category 1

2. Drywell and Torus Hydrogen Concentration Type A, Category 1

3. Reactor pressure Type 8, Category 1.

11

"The torus and drywell oxygen concentration is not used to initiate a safety function or to key the operator to perform a manual action. Browns Ferry's primary containment is operated with an oxygen deficient (i.e.,

inerted) atmosphere as one part of those measures for combustible gas control. The Containment Atmospheric Dilution (CAD) system is used following a postulated loss-of-coolant accident (LOCA) to dilute the containment atmosphere with nitrogen to maintain the hydrogen and oxygen concentrations below combustible levels. Hydrogen concentration is used at Browns Ferry to alert the operators to manually initiate the CAD System (H2 concentration is a Type A variable). The 02 concentration is used only as a surveillance instrument. NRC approved this configuration for units 1, 2, and 3 in technical specification amendments 38, 36, and 12 respectively."

1 We find this devisiation acceptable.

3.3.11 Containment Effluent Radioactivit Effluent Radioactivit Regulatory Guide 1.97 recommends the following instrumentation for noble gas. For containment effluent radioactivity, Category 3 instrumentation with a range from 10 to 10 uCi/cc. The licensee has "committed to installing a system to monitor the Browns Ferry stack For high-range noble gas with particulate and iodine collection on appropriate collection media in response to NUREG-0737, item II.F.l.l. and II.F.1.2.

"The Browns Ferry plant is designed to have one designated release point; namely, the stack. The secondary containment features of the plant will isolate and/or realign to cleanup systems which exhaust to the designated release point. Therefore, there is a very low probability of a major release of activity within other plant zones such as the turbine building. If an accidental release does occur in other areas, a high-radiation alarm is received and the effluent vent dampers and fans can be quickly isolated.

Since release paths such as the turbine building vents do not have cleanup systems, the isolation and/or shutdown of these ventilation system exhausts are stopped, it is not possible to determine quantitative releases."

12

The licensee should provide the information required in Section 6.2 of NUREG-0737, Supplement 1, identify any deviation from the recommendations of Regulatory Guide 1.97 and provide justification for any deviation.

3.3.12 Radiation Ex osure Rate Regulatory Guide 1.97 recommends instrumentation for this variable with a range of 10 to 10 R/hr. The instrumentation supplied by the licensee for this variable has a range of 10 to 10 mR/hr (10 to 10 mR/hr in the TIP room).

The licensee justifies the range by stating that, "in general, access is not required to any area of the secondary containment in order to service equipment important to safety in a post-accident situation. If and when accessibility is re-established in the long term, it will be done by a combination of portable radiation survey instruments and post-accident sampling of the secondary containment atmosphere. The existing lower range area radiation monitors would be used only in those instances in which radiation level were very mild."

Access to equipment areas could be required after an accident even if the areas are not designed for equipment service during an accident. The licensee does not provide enough justification for the deviation from the recommended range.

The licensee should identify the ranges and show them to be adequate, for the portable radiation monitors and their alternate instrumentation for long term surveillance and release assessment.

3.3.13 Su ression Chamber S ra Flow Or ell S ra Flow Regulatory Guide 1.97 recommends Category 2 instrumentation with a range from 0 to 110 percent of design flow for these variables. The licensee does not provide a direct measure for these variables. The licensee states the following. "The drywell sprays can be used to control the pressure and 13

temperature of the drywell. Likewise, the suppression pool sprays can be used to control the pressure and temperature in the torus. The flow to the sprays i's monitored by a flow element which is common to both the drywell spray flow and the suppression pool spray flow. This flow element is monitored as variable D16, LPCI flow. The operator can determine that the indicated flow is the flow that is being diverted to the sprays by observing the position (in the main control room) of the valves in the RHR line. The effectiveness of these flows can be verified by pressure and temperature changes of the drywell and the torus. The drywell pressure and drywell temperature instrumentation have been classified as Category l."

Variable D16 (the licensee's designation), LPCI flow, has Category 3 instrumentation suppl'ied (see Section 3.3.1). The temperature of the suppression pool is indicated on Category 1 instrumentation. However, the licensee should upgrade the LPCI flow instrumentation to Category 2 and verify that the suppression pool pressure instrumentation is Category 2 instrumentation.

3.3.14 Dr ell Atmos here Tem erature Regulatory Guide 1.97 recommends instrumentation for this variable with a range of 40 to 440'F. The licensee has provided instrumentation for this variable with a range of 0 to 400'F. They did not supply justification for not monitoring from 400 to 440'F.

However, our examination of the Final Safety Analysis Report (Reference 6), Figure 14.6-11, shows that the maximum post-accident drywell temperature is less than 300'F.

Therefore, the range of 0 to 400'F is adequate, and this deviation is acceptable.

3.3.15 Primar S stem Safet Relief Valve Position Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable with indication of either closed-not closed or flow through or

pressure (0 to 50 psig) in the valve discharge lines. The licensee monitors this variable with two diverse methods using Category 3 instrumentation.

First, there .is flow indication by acoustic monitors. Second, there is temperature indication of 0 to 600'F for the valve discharge lines.

The licensee states that "this instrumentation was provided in response to NUREG-0737, item II.0.3. The two instrumentation systems provide highly reliable indication of the S/RV position which is used primarily for diagnostic purposes."

Three clarifications from this position of NUREG-0737 are (a) if the position indication is not safety grade, a reliable single channel direct indication powered from a vital instrument bus may be provided if backup methods of determining- valve position are available and are discussed in the emergency procedures, (b) the valve position indication should be seismically qualified consistent with the component. of system to which it is attached and (c) the position indication should be qualified for its appropriate environment (any transient or accident which could cause the relief or safety valve to lift).

We find this to be a good faith attempt, as defined in NUREG-0737, Supplement No. 1, Section 3.7 (Reference 3), to meet NRC requirements and is, therefore, acceptable. However, the licensee should provide the information required by Section 6.2 of Reference 3 for this variable.

3.3.16 Standb Li uid Control S stem SLCS Flow The licensee states the following. "The SLCS flow, as discussed in Regulatory Guide 1.97, is to monitor the operation of the (SLCS). The SLCS tank level is monitored in the control room along with pump operation. When the squib valves are opened and SLCS pumps started, the total contents of the tank are pumped to the reactor. Flow through the line to the reactor is indicated by an annunciator and a white light. The actual amount of flow to the reactor is not relevant since the total volume of the tank is to be pumped. SLCS operation is adequately monitored by (a) tank level decreasing, 15

(b) pump operability indication, and (c) neutron flux response. The indication of the amount of flow is not necessary to ensure system operation."

Based on the above justification, we find that the licensee's position meets the intent of Regulatory Guide 1.97 for this variable.

3.3.17 Hi h Radioactivit Li uid Tank Level The licensee justifies not having instrumentation for this variable as follows. "The radioactive waste systems are designed to dispose of the radioactive process wastes generated during plant operation. The system is designed to prevent the inadvertent release of significant quantities of radioactive material from the restricted area of the plant so that resulting 1

exposures are within the guideline values of 10 CFR 20. The radwaste facility is located in the Radwaste Building. The Radwaste Building has been designed wo withstand a design basis earthquake (OBE). Should the floor drain collector tank fail or overfill before isolation," (Note: The FSAR indicates that the lines discharging to the floor drain collector tank of an accident signal.) "the spilled liquid would be retained in the building. Because the leaks or spills from the radwaste system are retained within the radwaste building and have little or no effect on the site boundary dose rate and the radwaste system is not required after a OBA, the level of the floor drain collector tank is not required to monitor the operation of the system."

Based on the licensee's justification, we find that monitoring this variable in the Browns Ferry control room is not necessary.

3.3.18 Reactor Buildin or Secondar Containment Area Radiation The licensee states that this variable need not be implemented. The licensee reports that the use of local radiation exposure rate monitors to detect breach or leakage through primary containment penetrations results in ambiguous indications. This is due to the radioactivity in the primary containment, the radioactivity in the fluids flowing in emergency core coolant system piping and the amount and location of fluid and electrical penetrations. The licensee concludes that the use of the stack noble gas 16

effluent monitors is the proper way to accomplish the purpose of this variable.

For the Browns Ferry Mark I containment, the range is recommended to be 10 to 10 R/hr. The licensee has not shown how the range requirement is related to and satisfied by the stack noble gas effluent monitors.

The licensee should supply additional justification for not implementing this variable.

3.3.19 Noble Gas and Vent Flow Rate Common Plant Vent Regulatory Guide 1.97 recommends instrumentation for this variable with a range from 10 to 10 'uCi/cc. The licensee states that they are committed to install a system to monitor the noble gas (high range) for the plant stack, which is the common plant release point. While the licensee has not identified the range of the instrumentation, they do state that it satisfies the requirements of NUREG-0737, items II.F.1.1 and II.F.1.2. These requirements do not establish a minimum range. The maximum range specified encompasses the regulatory guide recommendation.

We find this to be a good faith attempt, as defined in NUREG-0737, Supplement No. 1, Section 3.7 (Reference 3), to meet NRC requirements and is, therefore, acceptable. However, the licensee should provide the information required by Section 6.2 of Reference 3 for this variable.

3.3.20 Particulates and Halo ens Regulatory Guide 1.97 recommends instrumentation for this variable with a range of 10 to 10 uCi/cc. The licensee is installing instrumentation.

They do state that this instrumentation satisfies the requirements of NUREG-0737, items II.F.1.1. and II.F.1.2. These requirements do not establish a minimum range. The maximum range specified encompasses the regulatory guide requirements.

17

We find this to be a good faith attempt, as defined in NUREG-0737, Supplement No. 1, Section 3.7 (Reference 3), to meet NRC requirements and is, therefore, aCceptable. However, the licensee should provide the information required by Section 6.2 of Reference 3 for this variable.

3.3.20 Particulates and Halo ens Regulatory Guide 1.97 recommends instrumentation for this variable with a range of 10 to 10 uCi/cc. The licensee is installing instrumentation for this variable; however, they have not identified the range of the instrumentation. They do state that this instrumentation satisfies the requirements of NUREG-0737, items II.F.1.1 and II.F.1.2. These requirements do not establish a mihimum range. The maximum range specified encompasses the regulatory guide requirements.

We find this to be a good faith attempt, as defined in KUREG-0737, Supplement No. 1, Section 3.7 (Reference 3), to meet NRC requirements and is, therefore, acceptable. However, the licensee should provide the information required by Section 6.2 of Reference 3 for this variable.

3.3.21 Airborne Radiohalo ens and Particulates Regulatory Guide 1.97 recommends portable sampling with onsite analysis for this variable with a range of 10 to 10 uCi/cc. The licensee is meeting this recommendation by laboratory analysis in accordance with NUREG-0737, item II.B.3. This item states that the range "shall be adequate to provide pertinent data to the operator in order to describe the radiological and chemical status of the reactor coolant system."

As the licensee and NUREG-0737 do not state the range for this variable, we are unable to determine the adequacy of the licensee's equipment.

Additionally, the description given by the licensee leads us to believe that this is part of the post-accident sampling facility, not portable sampling as recommended by Regulatory Guide 1.97.

18

We conclude that the licensee should provide additional description for the instrumentation utilized for this variable.

3.3.22 Plant and Environs Radiation Regulatory Guide 1.97 recommends portable instrumentation for this variable, with ranges of 10 to 10 R/hr photons and 10 to 10" rads/hr beta radiation and low-energy photons. The licensee is supplying portable instrumentation for this variable; however, they have not identified the

• instrument range as required by Section 6.2 of Reference 3.

We conclude that the licensee should supply this information, identify any deviations from the recommendations of the regulatory guide and justify any deviation.

3.3.23 Accident Sam lin Primar Coolant Containment Air and Sum The licensee's post-accident sampling facility provides sampling and analysis. However, there are deviations from the following recommendations.

l. The sumps are not sampled

2. The licensee's submittal does not show compliance with the range recommendations

3. The containment air is not analyzed for hydrogen or oxygen content (continuous online monitors are used instead).

The licensee takes exception to the guidance of Regulatory Guide 1.97 with respect to post-accident sampling capability. This exception goes beyond the scope of this review and is being addressed by the NRC as part of their review of HUREG-0737, Item 11.8.3.

19

4. CONCLUSIONS Based os our review, we find that the licensee either conforms to, or is justified in deviating from, the guidance of Regulatory Guide 1.97 with the following exceptions:

1. For the fourteen variables identified in Section 3.3.1, the licensee should identify the specific exceptions from the Category 2 recommendations and provide justification for each exception.

2. Neutron flux the licensee's present instrumentation is acceptable on an interim basis until Category 1 instrumentation is developed and installhd (Section 3.3.2).

3.

Coolant level in reactor the licensee shoul'd either justify not having level instrumentation for the lower 1/3 of the core height or provide the recommended instrument range (Section 3.3.3).

4. Containment isolation valve position there are seven valves for which the licensee should show that they are known to be closed; there are twenty valves for which the licensee should provide the recommended instrumentation (Section 3.3.7).

5. Primary containment area radiation-high range the licensee should identify the specific deviations for this instrumentation, and provide justification for these deviations (Section 3.3.9).

6. Containment effluent radioactivity the licensee should provide the information by Section 6.2 of Reference 3, identify any deviation from the recommendations of Regulatory Guide 1.97 and provide satisfactory justification for any deviation (Section 3.3.11).

20

7. Effluent radioactivity the licensee should provide the information required by Section 6.2 of Reference 3, identify any deviation from the recommendations of Regulatory Guide 1.97 and provide satisfactory justification for any deviation (Section 3.3.11).

8. Radiation exposure rate the licensee should identify the ranges of the portable radiation monitors and their alternate instrumentation for long term surveillance and release assessment (Section 3.3.12).

9. Suppression chamber spray flow the licensee should verify that the alternate instrumentation used for this variable is either Category 1 or 2 (Section 3.3.13).

1

10. Drywell spray flow the licensee should upgrade the LPCI flow instrumentation (used as alternate instrumentation for this variable) to Category 2 (Section 3.3.13).

11. Primary system safety relief valve position the licensee should provide the information required by Section 6.2 of Reference 3 (Section 3.3.15).

12. Reactor building or secondary containment area radiation the licensee should provide additional justification for not implementing this variable (Section 3.3.18).

13. Noble gas and vent flow rate-common plant vent the licensee should provide the information required by Section 6.2 of Reference 3 (Section 3.3.19).

14. Particulates and halogens the licensee should provide the information required by Section 6.2 of Reference 3 (Section 3.3.20).

15. Airborne radiohalogens and particulates the licensee should provide additional information on the instrumentation supplied for this variable (Section 3.3.21).

21

16. Plant and environ radiation the licensee should identify the range of the instrumentation for this variable; if any deviations from the recommendations of the regulatory guide exist, they should be identified and justified (Section 3.3.22).

22

5. REFERENCES

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

2. Instrumentation for Li ht-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions Durin and Followin an Accident, Regulatory Guide .97, Revision 2, U.S. Nuclear Regulatory Commission (NRC), Office of Standards Development, December 1980.

3. Clarification of THI Action Plan Re uirements Re uirements For Emer enc Res onse Ca abilit , NUREG-0737, Supplement No. , NRC, Office of Nuclear Reactor Regulat on, January 1983.

4. Tennessee Valley Authority letter, L. H. Mills to H. R. Denton, NRC, April 30, 1984. i

5. Instrumentation for Li ht-Water-Cooled Nuclear ower Plants to Assess Plant Environs Conditions Durin and Followin an Accident, Regulatory Guide .97, Revision 3, NRC, Office of Nuclear Regulatory Research, May 1983.

38272 23